ML15203A042

IR 05000454/2015008; 05000455/2015008, on 4/20/2015 - 6/16/2015; Byron Station, Units 1 and 2; Component Design Bases Inspection. (Nfa)

ML15203A042
Person / Time
Site:	Byron
Issue date:	07/21/2015
From:	Christine Lipa NRC/RGN-III/DRS/EB2
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
References
IR 2015008
Download: ML15203A042 (65)

See also: IR 05000454/2015008

Text

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION III

2443 WARRENVILLE RD. SUITE 210

LISLE, IL 60532-4352

July 21, 2015

Mr. Bryan C. Hanson

Senior VP, Exelon Generation Company, LLC

President and CNO, Exelon Nuclear

4300 Winfield Road

Warrenville, IL 60555

SUBJECT: BYRON STATION, UNITS 1 AND 2 - NRC COMPONENT DESIGN BASES

INSPECTION; INSPECTION REPORT 05000454/2015008; 05000455/2015008

AND NOTICE OF VIOLATION

Dear Mr. Hanson:

On June 16, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed a Component

Design Bases Inspection at your Byron Station, Units 1 and 2. The purpose of this inspection

was to verify that design bases have been correctly implemented for the selected risk-significant

components, and that operating procedures and operator actions are consistent with design and

licensing bases. The enclosed report documents the results of this inspection, which were

discussed on June 16, 2015, with Mr. B. Currier, and other members of your staff.

This inspection examined activities conducted under your license as they relate to public

health and safety to confirm compliance with the Commissions rules and regulations, and

with the conditions in your license. Within these areas, the inspection consisted of a selected

examination of procedures and representative records, field observations, and interviews with

personnel.

Based on the results of this inspection, the NRC has identified an issue that was evaluated

under the risk Significance Determination Process as having very-low safety significance

(Green). The NRC has also determined that a violation is associated with this issue. This

violation was evaluated in accordance with the NRC Enforcement Policy. The current

Enforcement Policy is included on the NRCs web site at http://www.nrc.gov/about-nrc/

regulatory/enforcement/enforce-pol.html.

B. Hanson

-2-

The violation is cited in the enclosed Notice of Violation (Notice), and the circumstances

surrounding it are described in detail in the subject inspection report. The violation is being

cited in the Notice because Byron Station, Units 1 and 2, failed to restore compliance and failed

to have objective plans to restore compliance in a reasonable period following the NRC

identification of an associated Non-Cited Violation (NCV) on June 15, 2012. The associated

NCV was documented in Inspection Report 05000454/2012007; 05000455/2012007.

You are required to respond to this letter, and should follow the instructions specified in the

enclosed Notice when preparing your response. If you have additional information that you

believe the NRC should consider, you may provide it in your response to the Notice. The NRC

review of your response to the Notice will also determine whether further enforcement action is

necessary to ensure compliance with regulatory requirements.

Based on the results of this inspection, the NRC has also determined that six additional

NRC-identified findings of very-low safety significance (Green) were identified. The findings

involved violations of NRC requirements. However, because of their very-low safety

significance, and because the issues were entered into your Corrective Action Program, the

NRC is treating the issues as NCVs in accordance with Section 2.3.2 of the NRC Enforcement

Policy. These NCVs are described in the subject inspection report.

If you contest the subject or severity of the Non-Cited-Violation, you should provide a response

within 30 days of the date of this inspection report, with the basis for your denial, to the

U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC

20555-0001, with copies to the Regional Administrator, Region III; the Director, Office of

Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the

NRC Resident Inspector at the Byron Station.

In addition, if you disagree with the cross-cutting aspect assigned to any finding in this report,

you should provide a response within 30 days of the date of this inspection report, with the basis

for your disagreement, to the Regional Administrator, Region III, and the NRC Resident

Inspector at the Byron Station.

B. Hanson

-3-

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public

Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy

of this letter, its enclosure, and your response (if any) will be available electronically for public

inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS)

component of the NRC's Agencywide Documents Access and Management System (ADAMS).

ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html

(the Public Electronic Reading Room).

Sincerely,

/RA/

Christine A. Lipa, Chief

Engineering Branch 2

Division of Reactor Safety

Docket Nos. 50-454; 50-455

License Nos. NPF-37; NPF-66

Enclosures:

(1) Notice of Violation

(2) IR 05000454/2015008; 05000455/2015008;

cc w/encl: Distribution via LISTSERV

NOTICE OF VIOLATION

Enclosure 1

Exelon Generation Company, LLC

Docket No. 50-454; 50-455

Byron Station, Units 1 and 2

License No. NPF-37; NPF-66

During an U.S. Nuclear Regulatory Commission (NRC) inspection conducted from

April 20, 2015, through May 22, 2015, a violation of NRC requirements was identified.

In accordance with the NRC Enforcement Policy, the violation is listed below:

Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI,

Corrective Action, states, in part, that measures shall be established to assure that

conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations,

defective material and equipment, and non-conformances are promptly identified and

corrected.

Contrary to the above, from June 15, 2012, to May 22, 2015, the licensee failed to

correct a condition adverse to quality (CAQ). Specifically, on June 15, 2012, the

NRC issued a Non-Cited Violation 05000454/2012007-05; 05000455/2012007-05 for the

failure to provide means to detect and isolate a leak in the emergency core cooling

system within 30 minutes for Byron Station, Units 1 and 2, as described in

Section 6.3.2.5 of the Updated Final Safety Analysis Report which is a CAQ. As of

May 22, 2015, the licensee had not corrected the CAQ in a reasonable time period.

Instead, the licensee created action tracking items to develop a plan to correct the

CAQ, and the associated due date was extended at least eight times.

This violation is associated with a Green Significance Determination Process finding.

Pursuant to the provisions of 10 CFR 2.201, Exelon Generation Company, LLC, is hereby

required to submit a written statement or explanation to the U.S. Nuclear Regulatory

Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to

the Regional Administrator, Region III; and the NRC Resident Inspector at the Byron Station,

Units 1 and 2, within 30 days of the date of the letter transmitting this Notice. This reply

should be clearly marked as a Reply to a Notice of Violation; VIO 05000454/2015008-09; 05000455/2015008-09, and should include for each violation: (1) the reason for the violation,

or, if contested, the basis for disputing the violation or severity level; (2) the corrective steps that

have been taken and the results achieved; (3) the corrective steps that will be taken; and (4) the

date when full compliance will be achieved. Your response may reference or include previous

docketed correspondence, if the correspondence adequately addresses the required response.

If an adequate reply is not received within the time specified in this Notice, an order or a

Demand for Information may be issued as to why the license should not be modified,

suspended, or revoked, or why such other action as may be proper should not be taken.

Where good cause is shown, consideration will be given to extending the response time.

If you contest this enforcement action, you should also provide a copy of your response, with

the basis for your denial, to the Director, Office of Enforcement, United States Nuclear

Regulatory Commission, Washington, DC 20555-0001.

2

Because your response will be made available electronically for public inspection in the

NRC Public Document Room or from ADAMS, accessible from the NRC Web site at

http://www.nrc.gov/reading-rm/adams.html, to the extent possible, it should not include any

personal privacy, proprietary, or safeguards information so that it can be made available to the

public without redaction. If personal privacy or proprietary information is necessary to provide

an acceptable response, then please provide a bracketed copy of your response that identifies

the information that should be protected and a redacted copy of your response that deletes

such information. If you request withholding of such material, you must specifically identify the

portions of your response that you seek to have withheld and provide in detail the bases for your

claim of withholding (e.g., explain why the disclosure of information will create an unwarranted

invasion of personal privacy or provide the information required by 10 CFR 2.390(b) to support

a request for withholding confidential commercial or financial information). If safeguards

information is necessary to provide an acceptable response, please provide the level of

protection described in 10 CFR 73.21.

In accordance with 10 CFR 19.11, you may be required to post this Notice within two working

days of receipt.

Dated this 21 day of July, 2015.

Enclosure 2

U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Docket No:

50-454; 50-455

License No:

NPF-37; NPF-66

Report No:

05000454/2015008; 05000455/2015008

Licensee:

Exelon Generation Company, LLC

Facility:

Byron Station, Units 1 and 2

Location:

Byron, IL

Dates:

April 20, 2015, through June 16, 2015

Inspectors:

N. Féliz Adorno, Senior Reactor Inspector, Lead

B. Palagi, Senior Operations Engineer

D. Betancourt Roldán, Reactor Inspector, Mechanical

M. Jones, Reactor Inspector, Mechanical

A. Greca, Electrical Contractor

J. Leivo, Electrical Contractor

Approved by:

Christine A. Lipa, Chief

Engineering Branch 2

Division of Reactor Safety

SUMMARY ................................................................................................................................ 2

REPORT DETAILS .................................................................................................................... 7

1. REACTOR SAFETY ....................................................................................................... 7

1R21 Component Design Bases Inspection (71111.21) ............................................... 7

4. OTHER ACTIVITIES .....................................................................................................29

4OA2 Identification and Resolution of Problems ..........................................................29

4OA6 Management Meetings ......................................................................................38

SUPPLEMENTAL INFORMATION ............................................................................................. 2

KEY POINTS OF CONTACT .............................................................................................. 2

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED ................................................... 2

LIST OF DOCUMENTS REVIEWED .................................................................................. 3

LIST OF ACRONYMS USED .............................................................................................19

2

SUMMARY

Inspection Report 05000454/2015008; 05000455/2015008, 4/20/2015 - 6/16/2015; Byron

Station, Units 1 and 2; Component Design Bases Inspection.

The inspection was a 3-week on-site baseline inspection that focused on the design of

components. The inspection was conducted by four regional engineering inspectors, and

two consultants. Seven Green findings were identified by the team. Six of these findings were

considered Non-Cited Violations of U.S. Nuclear Regulatory Commission (NRC) regulations

while one of these findings was considered a Notice of Violation of NRC regulations. The

significance of inspection findings is indicated by their color (i.e., greater than Green, or

Green, White, Yellow, Red), and determined using Inspection Manual Chapter (IMC) 0609,

Significance Determination Process, dated April 29, 2015. Cross-cutting aspects are

determined using IMC 0310, Aspects Within the Cross-Cutting Areas effective date

December 4, 2014. All violations of NRC requirements are dispositioned in accordance with

the NRCs Enforcement Policy, dated July 9, 2013. The NRC's program for overseeing the

safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor

Oversight Process, Revision 5, dated February 2014.

NRC-Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Green: The team identified a finding of very-low safety significance (Green), and an

associated cited violation of Title 10, Code of Federal Regulations (CFR), Part 50,

Appendix B, Criterion XVI, Corrective Actions, for the failure to correct a Condition

Adverse to Quality (CAQ). Specifically, on June 15, 2012, the U.S. Nuclear Regulatory

Commission (NRC) issued a Non-Cited Violation (NCV) for the failure to provide means

to detect and isolate a leak in the Emergency Core Cooling System (ECCS) within

30 minutes as described in the Updated Final Safety Analysis Report (UFSAR), which

is a CAQ. As of May 22, 2015, the licensee had not corrected the CAQ. This violation

is being cited because the licensee had not restored compliance, or demonstrated

objective evidence of plans to restore compliance in a reasonable period following the

identification of the CAQ. The licensee captured this finding into their Corrective Action

Program (CAP) to promptly restore compliance.

The performance deficiency was determined to be more than minor because it was

associated with the Mitigating Systems cornerstone attribute of procedure quality, and

affected the cornerstone objective of ensuring the availability, reliability, and capability of

mitigating systems to respond to initiating events to prevent undesirable consequences.

In addition, it was associated with the Barrier Integrity cornerstone attribute of procedure

quality, and affected the cornerstone objective of providing reasonable assurance that

physical design barriers protect the public from radionuclide releases caused by

accidents or events. The finding screened as very-low safety significance (Green)

because it did not result in the loss of operability or functionality, and it did not represent

an actual pathway in the physical integrity of reactor containment. Specifically, the

licensee reasonably demonstrated that an ECCS leak could be detected and isolated

before it could adversely affect long-term cooling of the plant. The team determined that

the associated finding had a cross-cutting aspect in the area of human performance

because the licensee did not use a consistent and systematic approach to make

decisions. Specifically, the creation and management of the associated corrective action

assignments were not consistent with the instructions contained in their CAP procedure.

[H.13] (Section 4OA2.1.b(1))

3

Severity Level IV. The team identified a Severity Level IV NCV of 10 CFR 50.59(d)(1),

Changes, Tests, and Experiments, and an associated finding of very-low safety

significance (Green) for the licensees failure to perform a written safety evaluation that

provided the bases for the determination that a change which resulted in the sharing of

the refueling water storage tanks (RWSTs) of both reactor units did not require a license

amendment. Specifically, the licensee did not evaluate the adverse effect of reducing

reactor unit independence. The licensee captured this issue into their CAP with a

proposed action to revise the associated calculation to remove the dependence on the

opposite unit, and/or review the implications of crediting the opposite unit RWST under

their 10 CFR 50.59 process.

The performance deficiency was more than minor because it was associated with the

Mitigating Systems cornerstone attribute of design control, and affected the cornerstone

objective of ensuring the availability, reliability, and capability of mitigating systems to

respond to initiating events to prevent undesirable consequences. In addition, it was

associated with the Barrier Integrity cornerstone attribute of design control, and affected

the cornerstone objective of providing reasonable assurance that physical design

barriers protect the public from radionuclide releases caused by accidents or events.

In addition, the associated traditional enforcement violation was more than minor

because the team could not reasonably determine that the changes would not have

ultimately required NRC prior approval. The finding screened as very-low safety

significance (Green) because it did not result in the loss of operability or functionality,

and it did not represent an actual open pathway in the physical integrity of the reactor

containment. Specifically, the licensee reviewed the affected calculation and reasonably

determined that enough conservatism existed such that adequate net positive suction

head (NPSH) could be maintained without sharing the RWSTs of both reactor units.

The team did not identify a cross-cutting aspect associated with this finding because it

was confirmed not to be reflective of current performance due to the age of the

performance deficiency. (Section 1R21.5.b(1))

Green. The team identified a finding of very-low safety significance (Green), and an

associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the

licensees failure to translate applicable design basis into Technical Specifications (TSs)

Surveillance Requirement 3.5.4.2 implementing procedures. Specifically, these

procedures did not verify the RWST vent line was free of ice blockage at the locations,

and during all applicable MODEs of reactor operation assumed by the ECCS and

containment spray (CS) pump NPSH calculation. The licensee captured this issue into

their CAP to reconcile the affected procedures and calculation.

The performance deficiency was determined to be more than minor because it was

associated with the Mitigating Systems cornerstone attribute of design control, and

affected the cornerstone objective of ensuring the availability, reliability, and capability of

mitigating systems to respond to initiating events to prevent undesirable consequences.

Additionally, it was associated with the Barrier Integrity cornerstone attribute of design

control, and affected the cornerstone objective of providing reasonable assurance that

physical design barriers protect the public from radionuclide releases caused by

accidents or events. The finding screened as very-low safety significance (Green)

because it did not result in the loss of operability or functionality, and it did not represent

an actual open pathway in the physical integrity of reactor containment. Specifically, the

licensee performed a historical review of the last 3 years of operation, and did not find

any instances in which the vent path temperature fell below 35 degrees Fahrenheit.

4

The inspectors did not identify a cross-cutting aspect associated with this finding

because it was confirmed not to be reflective of current performance due to the age

of the performance deficiency. (Section 1R21.5.b(2))

Severity Level IV. The team identified a Severity Level IV NCV of 10 CFR 50.59(d)(1),

Changes, Tests, and Experiments, and an associated finding of very-low safety

significance (Green) for the licensees failure to perform a written evaluation that

provided the bases for the determination that the changes to the emergency service

water cooling tower (SXCT) tornado analysis as described in the UFSAR did not require

a license amendment. Specifically, the associated 10 CFR 50.59 Evaluation did not

address the introduction of a new failure mode, the resulting loss of heat removal

capacity during worst postulated conditions, and addition of operator actions that have

not been demonstrated can be completed within the required time to restore the required

SXCT heat removal capacity during worst case conditions. The licensee captured this

issue in their CAP with a proposed action to revise the 10 CFR 50.59 Evaluation and

submit a Licensee Amendment Request.

The performance deficiency was determined to be more than minor because it was

associated with the Mitigating Systems cornerstone attribute of protection against

external events, and affected the cornerstone objective of ensuring the availability,

reliability, and capability of mitigating systems to respond to initiating events to prevent

undesirable consequences. In addition, the associated tradition enforcement violation

was determined to be more than minor because the team could not reasonably

determine that the changes would not have ultimately required prior NRC approval.

The finding screened as of very-low safety significance (Green) using a detailed

evaluation because a loss of SXCT during a tornado event would degrade one or more

trains of a system that supports a risk-significant system or function. The bounding

change to the core damage frequency was less than 5.4E-8/year. The team did not

identify a cross-cutting aspect associated with this finding because the finding was not

representative of current performance due to the age of the performance deficiency.

(Section 1R21.5.b(3))

Green. The team identified a finding of very-low safety significance and an associated

NCV of TS 5.4, Procedures, for the failure to maintain emergency operating

procedures (EOPs) for transfer to cold leg recirculation. Specifically, the EOPs for

transfer to cold leg recirculation did not contain instructions for transferring the ECCS

and CS systems to the recirculation mode that ensured prevention of potential pump

damage when the RWST is emptied. The licensee captured this finding into their CAP

to create a standing order instructing operators to secure all pumps aligned to the RWST

when it is emptied, and implement long term corrective actions to restore compliance.

The performance deficiency was determined to be more than minor because it was

associated with the Mitigating Systems cornerstone attribute of procedure quality, and

affected the cornerstone objective of ensuring the availability, reliability, and capability of

mitigating systems to respond to initiating events to prevent undesirable consequences.

In addition, it was associated with the Barrier Integrity cornerstone attribute of procedure

quality, and affected the cornerstone objective of providing reasonable assurance that

physical design barriers protect the public from radionuclide releases caused by

accidents or events. The finding screened as of very-low safety significance (Green)

because it did not result in the loss of operability or functionality of mitigating systems,

represent an actual open pathway in the physical integrity of reactor containment, and

5

involved an actual reduction in function of hydrogen igniters in the reactor containment.

Specifically, the incorrect caution would only be used in the event that transfer to sump

recirculation was not completed prior to reaching tank low-level, or if the RWST suction

isolation valves fail to close. With respect to transfer to sump recirculation prior to

reaching tank low-level, a review of simulator test results reasonably determined that

operators reliably complete the transfer to sump recirculation prior to reaching this set

point. With respect to the failure of the RWST suction isolation valves, a review of

quarterly test results reasonably determined the valves would have isolated the tank

when required. The team did not identify a cross-cutting aspect associated with this

finding because it was not confirmed to reflect current performance due to the age of the

performance deficiency. (Section 1R21.6.b(1))

Green. The team identified a finding of very-low safety significance (Green), and an

associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,

and Drawings, for the failure to make an operability determination without relying on the

use of probabilistic tools. Specifically, an operability evaluation for an SXCT degraded

condition used probabilities of occurrence of tornado events which was contrary to the

requirements of the licensee procedure established for assessing operability of

structures, systems, and components (SSCs). The licensee captured the teams

concern in their CAP to revise the affected operability evaluation without using

probability of occurrence of tornado events.

The performance deficiency was more than minor because it was associated with the

Mitigating Systems cornerstone attribute of protection against external events, and

affected the cornerstone objective of ensuring the availability, reliability, and capability of

mitigating systems to respond to initiating events to prevent undesirable consequences.

The finding screened as of very-low safety significance (Green) using a detailed

evaluation because a loss of SXCT during a tornado event would degrade one or more

trains of a system that supports a risk-significant system or function. The bounding

change to the core damage frequency was less than 5.4E-8/year. The team determined

that this finding had a cross-cutting aspect in the area of human performance because

the licensee did not ensure knowledge transfer to maintain a knowledgeable and

technically competent workforce. Specifically, the licensee did not ensure personnel

were trained on the prohibition of the use of probabilities of occurrence of an event

when performing operability evaluations, which was contained in licensee procedure

established for assessing operability of SSCs. [H.9] (Section 4OA2.1.b(3))

Cornerstone: Barrier Integrity

Green. The team identified a finding of very-low safety significance, and an associated

NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and

Drawings, for the failure to have procedures to maintain the accuracy within necessary

limits of the instrument loops used to verify compliance with the containment average

air temperature TS limit of 120 degrees Fahrenheit. Specifically, in 2007, the licensee

cancelled the periodic preventive maintenance (PM) intended to maintain the necessary

instrument loops accuracy. The licensee entered this issue into their CAP and

reasonably established that the 120 degrees Fahrenheit limit was not exceeded

by reviewing applicable historical records from 2002 to time of this inspection.

6

The performance deficiency was determined to be more than minor because it was

associated with the configuration control attribute of the Barrier Integrity Cornerstone,

and adversely affected the cornerstone objective to ensure that physical design barriers

protect the public from radionuclide releases caused by accidents or events. The finding

screened as very-low safety significance (Green) because it did not represent an actual

open pathway in the physical integrity of reactor containment or involved an actual

reduction in hydrogen igniter function. Specifically, the containment integrity remained

intact and the finding did not impact the hydrogen igniter function. The team determined

that this finding had a cross-cutting aspect in the area of problem identification and

resolution because the licensee did not identify issues completely and accurately in

accordance with the CAP. Specifically, on January 15, 2015, the licensee captured the

lack of periodic PM activities for the containment air temperature instrument loops in the

CAP. However, the licensee failed to completely and accurately identify the issue in that

it was not treated as a CAQ. As a consequence, no corrective actions were

implemented. [P.1] (Section 4OA2.1.b(2))

7

REPORT DETAILS

1.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R21 Component Design Bases Inspection (71111.21)

.1

Introduction

The objective of the Component Design Bases Inspection (CDBI) is to verify that design

bases have been correctly implemented for the selected risk-significant components,

and that operating procedures and operator actions are consistent with design and

licensing bases. As plants age, their design bases may be difficult to determine, and

an important design feature may be altered or disabled during a modification. The

Probabilistic Risk Assessment (PRA) model assumes the capability of safety systems

and components to perform their intended safety function successfully. This inspectable

area verifies aspects of the Initiating Events, Mitigating Systems, and Barrier Integrity

cornerstones for which there are no indicators to measure performance.

Specific documents reviewed during the inspection are listed in the Attachment to the

report.

.2

Inspection Sample Selection Process

The team used information contained in the licensees PRA and the Byron Station,

Units 1 and 2, Standardized Plant Analysis Risk (SPAR) Model to identify two scenarios

to use as the basis for component selection. The scenarios selected were a feed and

bleed of the reactor coolant system (RCS), and a loss of ultimate heat sink (UHS).

Based on these scenarios, a number of risk-significant components, including those

with Large Early Release Frequency (LERF) implications, were selected for the

inspection.

The team also used additional component information such as a margin assessment

in the selection process. This design margin assessment considered original design

margin reductions caused by design modification, power uprates, or reductions due to

degraded material condition. Equipment reliability issues were also considered in the

selection of components for detailed review. These included items such as performance

test results, significant corrective actions, repeated maintenance activities, Maintenance

Rule (a)(1) status, components requiring an operability evaluation, U.S. Nuclear

Regulatory Commission (NRC) resident inspector input of problem areas/equipment,

and system health reports. Consideration was also given to the uniqueness and

complexity of the design, operating experience, and the available defense in depth

margins. A summary of the reviews performed and the specific inspection findings

identified are included in the following sections of the report.

The team also identified procedures and modifications for review that were associated

with the selected components. In addition, the team selected operating experience

issues associated with the selected components.

8

This inspection constituted 16 samples (12 components, of which 3 had LERF

implications, and 4 operating experience) as defined in Inspection

Procedure 71111.21-05.

.3

Component Design

a.

Inspection Scope

The team reviewed the Updated Final Safety Analysis Report (UFSAR), Technical

Specification (TS), design basis documents, drawings, calculations and other available

design basis information, to determine the performance requirements of the selected

components. The team used applicable industry standards, such as the American

Society of Mechanical Engineers Code, and Institute of Electrical and Electronics

Engineers Standards, to evaluate acceptability of the systems design. The NRC

also evaluated licensee actions, if any, taken in response to NRC issued operating

experience, such as Information Notices (INs). The review verified that the selected

components would function as designed when required and support proper operation of

the associated systems. The attributes that were needed for a component to perform its

required function included process medium, energy sources, control systems, operator

actions, and heat removal. The attributes to verify that the component condition and

tested capability were consistent with the design bases and appropriate may have

included installed configuration, system operation, detailed design, system testing,

equipment and environmental qualification, equipment protection, component inputs

and outputs, operating experience, and component degradation.

For each of the components selected, the team reviewed the maintenance history, PM

activities, system health reports, operating experience-related information, vendor

manuals, electrical and mechanical drawings, and licensee corrective action documents.

Field walkdowns were conducted for all accessible components to assess material

condition, including age-related degradation, and to verify that the as-built condition was

consistent with the design. Other attributes reviewed are included as part of the scope

for each individual component.

The following 12 components (samples) were reviewed:

Safety Injection Pump (1SI01PB): The team reviewed analyses associated

with inadvertent safety injection (SI) actuation and hydraulic calculations to

assess the pump capability to provide its required accident mitigation function.

The reviewed hydraulic analyses included pump minimum required flow, runout

flow, flow capacity/balance, minimum required net positive suction head (NPSH),

and air entraining vortices. In addition, the team reviewed a sample of operating

procedures associated with pump operation under normal and accident

conditions to assess their consistency with applicable design basis analyses.

The team also reviewed test procedures and completed surveillance tests,

including quarterly and comprehensive in-service testing and flow balances,

to assess the associated acceptance criteria and test results. The team also

reviewed the supporting electrical calculations associated with performance of

the SI pump under design basis conditions. This included review of brake

horsepower requirements for the pump motor, performance under degraded

voltage conditions, and motor protection to assess the capability of the motor to

perform its safety function under design basis conditions. In addition, the team

9

reviewed voltage drop calculations to assess the availability of direct current (DC)

control voltage at the associated bus needed to operate the pump circuit breaker.

The team also performed a non-intrusive visual inspection of the component to

assess overall material condition, configuration, and potential vulnerabilities to

hazards. To assess operating trends and the licensees ability to evaluate and

correct problems, the team reviewed system health reports, selected corrective

action documents, and PM procedures and records.

Pressurizer Power-Operated Relief Valve (1RY456): The team reviewed the

pressure and temperature limit report and calculations associated with the

power-operated relief valve (PORV) lift settings, relief capacity, and set points for

low-temperature overpressure (LTOP) scenarios to assess the PORV capability

to provide its RCS overpressure protection function. The team also reviewed test

procedures and completed surveillances to assess the associated acceptance

criteria and test results. In addition, the team reviewed a sample of associated

operating procedures to assess their consistency with applicable design basis

analyses. The team also reviewed the schematic diagrams for the PORV control

circuit to assess its suitability for bleed-and-feed operation as prescribed by

operating procedures, and to assess the pilot solenoid and position limit switches

qualification for post-accident environmental conditions. The team reviewed

voltage drop calculations to assess the availability of the voltage needed at the

solenoid valve to operate the PORV. The team also reviewed control wiring

schematics and associated instrument loop diagrams to assess the consistency

between operations and system design requirements. This review included a

circuit protection evaluation intended to demonstrate that the containment

electrical penetration was not adversely affected by in-containment faults. The

team also reviewed documentation associated with environmental qualifications

for the postulated containment accident conditions and replacement of

components susceptible to aging. The team reviewed system health reports,

selected corrective action documents, and PM procedures and records to assess

operating trends and the licensees ability to evaluate and correct problems.

Power-Operated Relief Valve Accumulator (1RY32MB): The team reviewed the

accumulator sizing calculation, PORV pressure set point, accumulator stress

analysis, and maximum allowed accumulator leak rate to assess the accumulator

capability to supply the required amount of air pressure and volume to stroke

open its associated PORV on a loss of normal air supply. Additionally, the team

reviewed the design calculation that established the minimum number of PORV

strokes required during certain events, such as LTOP and natural circulation

cooldown. The team also reviewed test procedures and completed surveillances

to assess the associated acceptance criteria and test results. In addition, the

team reviewed a sample of associated operating procedures to assess their

consistency with applicable design basis analyses. Finally, the team reviewed

system health reports, selected corrective action documents, and recent

modifications and operability evaluations to assess operating trends and the

licensees ability to evaluate and correct problems.

Refueling Water Storage Tank (1SI01T): The team reviewed a sample of

associated operating procedures under normal and emergency conditions to

assess their consistency with applicable design basis analyses. The team

also performed a non-intrusive visual inspection of the refueling water storage

10

tank (RWST) to assess overall material condition, configuration, and potential

vulnerabilities to hazards. To assess operating trends, component health, and

the licensees ability to evaluate and correct problems, the team reviewed system

health reports, selected corrective action documents, and recent modifications.

The team reviewed design analyses associated with the ability of the RWST

system to maintain its design function during external events such as tornados

and earthquakes. Additionally, the team reviewed design calculations related to

level set points, temperature limits, and minimum required RWST volume to

mitigate a loss of coolant accident (LOCA), and to support feed-and-bleed

scenarios. The team also reviewed the schematic diagrams and instrument

uncertainty calculations to assess the low-low RWST level signal (i.e., LO-2)

capability to automatically open the containment sump isolation valves

(i.e., 1SI8811A/B) following a LOCA, and its consistency with the associated set

point calculation including instrument uncertainty considerations. To assess

operating trends, component health, and the licensees ability to evaluate and

correct problems, the team reviewed system health reports, selected corrective

action documents, recent modifications, and PM/calibration procedures and

records.

Emergency Service Water Makeup Pump (0SX02PA): The team reviewed

design documents and procedures to assess consistency with vendor

specifications. The team reviewed calculations associated with pump capability

and performance to assess the pump capability to perform its design function of

providing sufficient inventory to the associated Emergency Service Water

Cooling Tower (SXCT) basin under different postulated scenarios. The team

reviewed the water inventory availability from the suction source under routine

service as well as extreme conditions. This review included low and high-river

water levels and temperatures, pump NPSH, pump suction submergence, and

minimum flow protection. The team also reviewed procedures associated with

protection against flooding, seismic, and tornado events since the makeup pump

is credited to some extent during these postulated events. The team also

performed a non-intrusive visual inspection of the pump to assess overall

material condition, configuration, and potential vulnerabilities to hazards.

Work orders and maintenance procedures were reviewed to verify effectiveness

of site maintenance. The team also reviewed test procedures and completed

surveillances to assess the associated acceptance criteria and test results.

To assess operating trends, component health, and the licensees ability to

evaluate and correct problems, the team reviewed system health reports and

selected corrective action documents.

Emergency Service Water Makeup Pump Diesel Engine (0SX02PA-K): The

team reviewed design documents and procedures to assess consistency with

vendor specifications. The team reviewed diesel fuel oil day tank level alarm

response procedures and sizing analyses including the engine diesel fuel oil

consumption rate calculation, tank capacity, vortexing calculation, level

indicators, and alarm setpoint. In addition, the team reviewed the control circuit

electrical diagram to assess the consistency between operations and design

basis requirements. The team also reviewed the set point calculation for the

SXCT basin level switch associated with the starting logic of the diesel engine

to assess consistency between the specified setting and applicable design basis

requirements. In addition, the team reviewed recent level instrument calibration

11

results. The team also reviewed circuit protection and control voltage to assess

the diesel engine capability to start on demand. The inspectors reviewed

completed work orders to assess the as-found and as-left condition of the

diesel engine following recent maintenance activities. The team also reviewed

test procedures and completed surveillances to assess the associated

acceptance criteria and test results. The team also performed a non-intrusive

visual inspection of the engine to assess overall material condition, configuration,

and potential vulnerabilities to hazards. To assess operating trends and the

licensees ability to evaluate and correct problems, the team reviewed system

health reports, selected corrective action documents, modifications, and PM

procedures and records.

Emergency Service Water Cooling Tower (0SX02AA/B and 0SX03CA/H):

The team reviewed design calculations and procedures associated with fan

performance, basin sizing, heat transfer, and makeup requirements during

postulated events including LOCA, tornado, and seismic events. The electrical

calculations associated with fan performance under design basis conditions

were reviewed to assess consistency with the design bases and the motor

capability to perform its specified safety function. This review considered fan

motor brake horsepower requirements, performance under degraded voltage

conditions, and motor protection. The team reviewed voltage drop calculations

to assess the availability of the DC control voltage needed at the associated load

center for the closing and tripping of the cooling tower fan circuit breakers. The

team also reviewed the alternating current (AC) and DC electrical distribution

systems to assess the SXCT capability to perform its specified safety function

assuming a single failure of electrical components. The team also reviewed

control wiring diagrams of the deep well pump and associated control valves to

assess consistency between their operation and design requirements. The team

also performed a non-intrusive visual inspection of the SXCT basin structure, fan

motors, valve houses, and electrical equipment rooms to assess overall material

condition, configuration, and potential vulnerabilities to hazards. The team also

reviewed test procedures and completed surveillances to evaluate the associated

acceptance criteria and test results. To assess operating trends and the

licensees ability to evaluate and correct problems, the team reviewed system

health reports, selected corrective action documents, operability evaluations,

modifications, and PM procedures and records.

4160 Volts Alternating Current Bus 142: The team reviewed voltage drop

calculations to assess the availability of the DC control voltage needed at the

associated bus for the operation of the associated circuit breakers. The team

reviewed calculations associated with load flow, degraded voltage, and protective

settings for selected electrical load paths served by the bus and associated with

the inspection samples to assess the bus capability to support the loads required

safety functions under design basis conditions. The team also performed a

non-intrusive visual inspection of the switchgear to assess overall material

condition, configuration, and potential vulnerabilities to hazards or extreme

service environments. To assess operating trends and the licensees ability to

evaluate and correct problems, the team reviewed system health reports,

selected corrective action documents, and selected PM procedures and records.

12

120 Volts Alternating Current Instrument Bus 111: The team reviewed the DC

voltage drop calculations to assess the availability of the voltage needed for the

proper operation of the associated inverter, including during a loss of AC power.

The team also reviewed the bus loading and breaker ratings to assess the bus

and loads protection against spurious tripping. In addition, the team reviewed a

modification which installed forced air cooling units for the inverter serving the

bus to assess the modification implementation and any potential impact on the

inverter. To assess operating trends and the licensees ability to evaluate and

correct problems, the team reviewed system health reports, selected corrective

action documents, and PM procedures and records for the bus.

125 Volts Direct Current Bus 111: The team reviewed bus loading and short

circuit calculations as well as cable, bus, and circuit breaker ratings to assess

bus and cable capabilities of carrying the maximum anticipated loading and

protection against faulted conditions. The team also reviewed voltage drop and

battery sizing calculations to assess the capability to support momentary and

continuous loading for the duration of the duty cycle during accident conditions

and the loss of all AC power (i.e., station blackout). Additionally, the team

reviewed the battery charger sizing calculation to assess its capability of

maintaining the battery in a charged state and recharging the battery in a timely

manner following a loss of AC power event. The team also reviewed room

heat-up calculations to ensure that the DC components were not adversely

affected by steam line breaks in the turbine building. In addition, the team

reviewed purchase specifications, vendor documents, seismic test reports,

certificate of compliance, and cable separation to assess consistency of the

installed component to the design requirements. For the battery, this review

included an assessment of the inter-cell resistance conformance to voltage drop

calculations. Breaker/fuse coordination was also reviewed to assess the

capability to interrupt overloads and faulted conditions. The team also reviewed

testing procedures and associated recent results, recent system health reports,

molded-case circuit breaker testing, maintenance activities, and recent corrective

action documents to assess component health history.

24 Volts Direct Current Bus 035-2: The team reviewed the sizing calculation for

the diesel start system and the control batteries to assess their capability of

providing adequate voltage to the associated components for the duration of the

duty cycle during accident conditions and loss of all AC power. The team also

reviewed components and wiring schematics related to the diesel start and

control logic to assess the bus capability to perform its intended function.

Additionally, the team reviewed the battery charger sizing calculation to assess

its capability to maintain the batteries in a charged state, and to recharge them in

a timely manner following a loss of AC power event. The team reviewed

purchase specifications, vendor documents, seismic test report, and certificate of

conformance to assess consistency of the installed component to the design

requirements. The team also reviewed testing procedures and associated recent

results, health reports, maintenance activities, and recent corrective action

documents to assess component health history.

480 Volts Alternating Current Motor Control Center 132Z1: The team assessed

conformance to the applicable design and licensing basis by performing an

engineering review of the motor control center (MCC) loading, MCC and control

13

circuits degraded voltage and maximum voltage, electrical protection, and

electrical isolation/physical circuit separation of the MCC from non-safety class

loads. The loads considered during this review were the SXCT riser motor

operated valves (MOVs) (i.e., 0SX163E/F), SXCT makeup MOV (i.e., 0SX157A),

and basin bypass MOV (i.e., 0SX162B). The team reviewed the calculations that

determined minimum terminal voltages for these MOVs to assess consistency

with the associated MOV thrust calculations. The team also reviewed the

thermal overload sizing calculations for these MOV circuits to assess their

protection against premature thermal overload trip and the minimum voltage

calculations for the 120 volts alternating current (VAC) service to the SXCT basin

level control system to assess the availability of the voltage needed for the level

instrumentation under design basis conditions. To evaluate whether there were

adverse operating trends and to assess the licensees ability to evaluate and

correct problems, the team reviewed system health reports, selected corrective

action documents, and PM procedures and records for the MCC.

b.

Findings

(1) Question Regarding the Maximum Wet Bulb Temperature Value Assumed in the

Emergency Service Water Cooling Tower Tornado Analysis

Introduction: The team identified an unresolved item (URI) regarding the maximum

wet-bulb temperature value assumed in the SXCT tornado analysis. Specifically, the

team noted the analysis used a value which was less restrictive than the highest 3-hour

wet-bulb temperature recorded for the site as described in the UFSAR.

Description: In Section 3.5.4 of the UFSAR, Analysis of Missiles Generated by a

Tornado, stated that, An analysis of the UHS cooling capability for a tornado missile

event has been made. It also stated that, A maximum outside air wet-bulb temperature

of 78 degrees Fahrenheit is assumed and is conservatively held constant throughout the

transient. In addition, this UFSAR section stated that, The analysis was performed

using service water cooling tower performance curves generated using the method

described in UFSAR Section 9.2.5.3.1.1.2 [...]. The analysis of the UHS cooling

capability for a tornado missile event was calculation BYR09-002, UHS Capability with

Loss of SX [Emergency Service Water] Fans due to a Tornado Event, which used a

constant maximum outside air wet-bulb temperature value of 78 degrees Fahrenheit

consistent with UFSAR Section 3.5.4.

However, the team noted the assumed maximum outside air wet-bulb temperature value

of 78 degrees Fahrenheit appeared to be inconsistent with the method described in

UFSAR Section 9.2.5.3.1.1.2, Steady State Tower Performance Analysis. Specifically,

it stated that, The design wet-bulb temperature during warm weather operation is

82 degrees Fahrenheit (Refer to UFSAR Section 2.3.1.2.4). In Section 2.3.1.2.4 of

the UFSAR, Ultimate Heat Sink Design, stated that, This analysis [described in

Section 9.2.5.3.1.1] includes scenarios with the highest 3-hour wet-bulb temperature,

82 degrees Fahrenheit, which was recorded on July 30, 1961, at 3:00 pm. This UFSAR

section also stated that, Per Regulatory Guide 1.27, the ultimate heat sink must be

capable of performing its cooling function during the design basis event for this worst

case 3-hour wet-bulb temperature. In addition, it stated, However, the design

operating wet-bulb temperature of the ultimate heat sink is 78 degrees Fahrenheit

(ASHRAE 1 percent exceedance value).

14

This issue is unresolved pending further review by the Office of Nuclear Reactor

Regulation (NRR) of the licensing basis related to the wet-bulb temperature value

applicable for the SXCT tornado analysis, and the team determination of further NRC

actions to resolve the issue. (URI 05000454/2015008-01; 05000455/2015008-01,

Question Regarding the Maximum Wet-Bulb Temperature Value Assumed in the SXCT

Tornado Analysis)

(2) Maximum Wet-Bulb Temperature Value Assumed in Emergency Service Water Cooling

Tower Analysis Was Not Monitored

Introduction: The team identified an URI regarding the lack of monitoring the maximum

wet-bulb temperature value assumed in SXCT analysis. Specifically, the team noted the

maximum wet-bulb temperature value was a critical parameter for the SXCT analyses,

but the licensee had not established a testing program to verify actual values were

bounded.

Description: In Section 3.5.4 of the UFSAR, Analysis of Missiles Generated by a

Tornado, stated that, An analysis of the UHS cooling capability for a tornado missile

event has been made. It also stated that, A maximum outside air wet-bulb temperature

of 78 degrees Fahrenheit is assumed, and is conservatively held constant throughout

the transient.

In Section 9.2.5.3.1.1 of the UFSAR, Design Basis Reconstitution, stated that,

The design basis event for the Byron ultimate heat sink is a LOCA coincident with a

loss-of-off-site power (LOOP) in one unit, and the concurrent orderly shutdown from

maximum power to cold shutdown of the other unit using normal shutdown operating

procedures. It also stated that, The design wet-bulb temperature during warm

weather operation is 82 degrees Fahrenheit (Refer to the UFSAR Section 2.3.1.2.4).

In Section 2.3.1.2.4 of the UFSAR, Ultimate Heat Sink Design, stated that, This

analysis [described in Section 9.2.5.3.1.1] includes scenarios with the highest 3-hour

wet-bulb temperature, 82 degrees Fahrenheit, which was recorded on July 30, 1961, at

3:00 pm.

The analysis of the UHS cooling capability for a tornado missile event was calculation

BYR09-002, UHS Capability with Loss of SX Fans due to a Tornado Event, which used

a constant maximum outside air wet-bulb temperature value of 78 degrees Fahrenheit

consistent with UFSAR Section 3.5.4. The analysis of the UHS cooling capability for a

LOCA coincident with a LOOP was calculation UHS-01, Ultimate Heat Sink Design

Basis LOCA Single Failure Scenarios, which used a constant maximum outside air

wet-bulb temperature value of 82 degrees Fahrenheit consistent with the UFSAR

Section 9.2.5.3.1.1.

However, the licensee had not established a testing program to verify actual

environmental conditions were bounded by these analyses and design basis limits.

In response to the team questions, the licensee stated that this approach was

acceptable because historical data showed wet-bulb temperature had a cyclic nature,

maximum wet-bulb temperature lasted for relatively short durations, and the analyses

assumed constant wet-bulb temperature values.

15

This issue is unresolved pending further NRR review of the acceptability of the

licensee approach to ensure the SXCT analyses bounded actual environmental

conditions, and the team determination of further NRC actions to resolve the issue.

(URI 05000454/2015008-02; 05000455/2015008-02, Maximum Wet-Bulb Temperature

Value Assumed in SXCT Analysis Was Not Monitored)

.4

Operating Experience

a.

Inspection Scope

The team reviewed four operating experience issues (samples) to ensure that NRC

generic concerns had been adequately evaluated and addressed by the licensee.

The operating experience issues listed below were reviewed as part of this inspection:

IN 2013-05, Battery Expected Life and Its Potential Impact on Surveillance

Requirements;

IN 2010-26, Submerged Electrical Cables;

IN 2013-12, Improperly Sloped Instrument Sensing Lines; and

IN 2012-01, Refueling Water Storage Tank Degradation.

b.

Findings

No findings were identified.

.5

Modifications

a.

Inspection Scope

The team reviewed five permanent plant modifications related to selected risk-significant

components to verify that the design bases, licensing bases, and performance capability

of the components had not been degraded through modifications. The modifications

listed below were reviewed as part of this inspection effort:

Engineering Change (EC) 385951, Multiple Spurious Operation - Scenario 14,

1SI8811A/B;

EC396016, Increase U1 Pressurizer PORV Accumulator Tank Operating

Pressure to Increase Number of PORV Open/Close Cycles from Accumulator;

EC388735, Detailed Review of the FC Purification for Use of Non-Safety

Related Portion Connected to Safety Related Piping;

DRP 11-052, Clarify References to RWST Internal Pressure in the ECCS and

the CS Pumps NPHS Analysis; and

EC385829, Tornado Missile Design Basis for the Essential Service Water

Cooling Towers.

16

b.

Findings

(1) Failure to Evaluate the Adverse Effects of Sharing the Refueling Water Storage Tanks of

Both Reactor Units

Introduction: The team identified a Severity Level IV NCV of 10 CFR 50.59(d)(1),

Changes, Tests, and Experiments, and an associated finding of very-low safety

significance (Green) for the licensees failure to perform a written safety evaluation that

provided the bases for the determination that a change which resulted in the sharing of

the RWSTs of both reactor units did not require a license amendment. Specifically,

screening 6E-05-0172, UFSAR Change Package (DRP)11-052, did not address the

reduction in reactor unit independence associated with sharing the RWSTs air space of

both reactor units.

Description: Each reactor unit has one RWST, which supplies borated water to both

trains of the Emergency Core Cooling System (ECCS) and Containment Spray (CS)

systems during the injection phase of a LOCA recovery. The UFSAR Section 6.3,

Emergency Core Cooling System, and UFSAR Section 6.5.2, Containment Spray

Systems, described the NPSH analyses for the ECCS and CS pumps when their

suctions are aligned to their associated RWST. Before November 16, 2005, these

UFSAR sections described the RWST as being under atmospheric pressure during

the injection mode. The licensee changed these UFSAR descriptions following the

discovery that the RWST would not be under atmospheric pressure because the RWST

vent did not have the capacity to prevent vacuum during the high outflow expected

during the injection phase, and the vent vacuum relief device was not safety related.

This discovery was captured in the CAP as AR00239280.

The licensee reviewed this UFSAR change in Title 10, Code of Federal Regulations

(CFR), Part 50.59 screening 6E-05-0172, Clarify References to RWST Internal

Pressure in the ECCS and CS Pumps NPSH Analysis. The screening concluded that

the change did not require a 10 CFR 50.59 safety evaluation and, consequently, NRC

prior approval because the change did not result in an adverse effect to the ECCS and

CS systems. Specifically, the licensee determined the expected vacuum would not

affect the structural integrity of the tank. In addition, the licensee determined in

calculation BYR 04-016, [Residual Heat Removal] RHR, SI, [Chemical and Volume

Control] CV, and CS Pump NPSH during ECCS Injection Mode, that the available

NPSH for the pumps while taking suction from the RWST remained adequate when

considering the expected vacuum.

However, the team noted that revised calculation BYR 04-016 credited the entire RWST

vent line, which was common to the RWSTs of both reactor units. Consequently, the

change credited the free air space of both tanks to mitigate the vacuum expected during

tank drawdown. The team also noted that UFSAR Section 3.1.2.1.5, Evaluation Against

Criterion 5 - Sharing of Structures, Systems, and Components, described those SSCs

important to safety shared by the two reactor units, and the RWSTs were not included

as shared SSCs. Thus, the team noted the licensee implemented a change to the

facility as described in the UFSAR that resulted in a reduction of reactor unit

independence. Changes to the facility as described in the UFSAR that reduce reactor

unit independence adversely impact 10 CFR 50.59 change evaluation criteria because

they result in more than a minimal increase in the likelihood of occurrence of a

malfunction of an SSC important to safety. Since the licensee failed to appropriately

17

evaluate this adverse effect in a 10 CFR 50.59 safety evaluation, the team could not

reasonably determine that the change would not have ultimately required NRC prior

approval.

The licensee captured this issue in their CAP as AR 02496142. The corrective actions

considered at the time of this inspection were to revise calculation BYR04-016 to not

credit the opposite unit RWSTs air space and/or revise 10 CFR 50.59 screening

6E-05-0172 to consider the implications of crediting the opposite unit RWST air space.

The team also noted the licensee did not correctly implement this change into

associated surveillance procedures intended to verify RWST operability. This separate

concern is discussed in detail in Section 1R21.5.b(2) of this report.

Analysis: The team determined that the failure to provide a written evaluation that

provided the bases for the determination that a change which resulted in the sharing of

the RWSTs of both reactor units did not require a license amendment, was contrary to

the requirements of 10 CFR 50.59(d)(1), and was a performance deficiency. The

performance deficiency was more than minor because it was associated with the

Mitigating Systems cornerstone attribute of design control, and affected the cornerstone

objective of ensuring the availability, reliability, and capability of mitigating systems to

respond to initiating events to prevent undesirable consequences. In addition, it was

associated with the Barrier Integrity cornerstone attribute of design control, and affected

the cornerstone objective of providing reasonable assurance that physical design

barriers protect the public from radionuclide releases caused by accidents or events.

Specifically, the change did not ensure the RWST capability to support ECCS and CS

mitigating and barrier functions because it eliminated the capability to achieve the RWST

supporting function while maintaining separation of the reactor units.

In addition, the associated violation was determined to be more than minor because the

team could not reasonably determine the changes would not have ultimately required

NRC prior approval.

Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process

instead of the Significance Determination Process (SDP) because they are considered

to be violations that potentially impede or impact the regulatory process. This violation is

associated with a finding that has been evaluated by the SD, and communicated with an

SDP color reflective of the safety impact of the deficient licensee performance. The

SDP, however, does not specifically consider the regulatory process impact. Thus,

although related to a common regulatory concern, it is necessary to address the violation

and finding using different processes to correctly reflect both the regulatory importance

of the violation and the safety significance of the associated finding.

In this case, the team determined that the finding could be evaluated using the SDP in

accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination

Process by using Attachment 0609.04, Initial Characterization of Findings. Since the

finding impacted the Mitigating Systems and Barrier Integrity cornerstones, the

inspectors screened the finding through IMC 0609 Appendix A, The Significance

Determination Process for Findings At-Power, using Exhibit 2, Mitigating Systems

Screening Questions, and Exhibit 3, Barrier Integrity Screening Questions. The

finding screened as very-low safety significance (Green) because it did not result in the

loss of operability or functionality, and it did not represent an actual open pathway in the

physical integrity of the reactor containment. Specifically, the licensee reviewed

18

calculation BYR 04-016, and reasonably determined that enough conservatism existed

such that adequate NPSH could be maintained without sharing the RWSTs of both

reactor units.

In accordance with Section 6.1.d of the NRC Enforcement Policy, this violation is

categorized as Severity Level IV because the resulting change was evaluated by the

SDP as having very-low safety significance (i.e., Green finding).

The inspectors did not identify a cross-cutting aspect associated with this finding

because it was confirmed not to be reflective of current performance. Specifically, the

finding occurred approximately 10 years ago.

Enforcement: Title 10 CFR 50.59, Changes, Tests, and Experiments, Section (d)(1)

requires, in part, the licensee to maintain records of changes in the facility, of changes in

procedures, and of tests and experiments made pursuant 10 CFR 50.59(c). These

records must include a written evaluation which provides the bases for the determination

that the change, test, or experiment does not require a license amendment pursuant to

Paragraph (c)(2) of this section. Paragraph (c)(2)(ii) states, in part, that a licensee shall

obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed

change, test, or experiment if the change, test, or experiment would result in more than a

minimal increase in the likelihood of occurrence of a malfunction of an SSC important to

safety previously evaluated in the UFSAR. In the UFSAR Sections 6.3 and 6.5 describe

the NPSH evaluations for ECCS and CS pumps when their suctions are aligned to their

associated RWST. Additionally, UFSAR Section 3.1.2.1.5 states that Those systems,

structures, and components important to safety shared by the two units are the ultimate

heat sinks and the associated Byron makeup water systems; various heating, ventilating,

and air conditioning systems within the shared auxiliary and fuel handling building; and a

component cooling heat exchanger which can be valved to serve one unit or the other.

The RWSTs are not included as shared SSCs.

Contrary to the above, on November 16, 2005, the licensee failed to maintain a record

of a change in the facility made pursuant to 10 CFR 50.59(c) that included a written

evaluation which provided the bases for the determination that the change did not

require a license amendment pursuant to 10 CFR 50.90(c)(2). Specifically, the licensee

changed the ECCS and CS pumps NPSH calculation for their injection mode of

operation (i.e., calculation BYR 04-016) to credit the entire vent line common to the

RWSTs of both reactor units and, consequently, the free air space of both tanks to

mitigate the vacuum expected during tank drawdown. However, the licensee failed to

perform a written evaluation that provided the bases for the determination that the

change effect of reducing reactor unit independence by sharing their RWSTs did not

result in more than a minimal increase in the likelihood of occurrence of a malfunction of

the RWSTs and their supported safety systems.

The licensee is still evaluating its planned corrective actions. However, the team

determined that the continued non-compliance does not present an immediate safety

concern because the licensee reasonably determined that the affected analysis

contained enough conservatism such that adequate NPSH could be maintained

without sharing the RWSTs of both reactor units.

19

Because this was a Severity Level IV violation and was entered into the licensee

Corrective Action Program (CAP) as AR 02496142, this violation is being treated

as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy.

(NCV 05000454/2015008-03; 05000455/2015008-03; Failure to Evaluate the Adverse

Effects of Sharing the RWSTs of Both Reactor Units)

The associated finding is evaluated separately from the traditional enforcement violation

and, therefore, the finding is being assigned a separate tracking number.

(FIN 05000454/2015008-04; 05000455/2015008-04; Failure to Evaluate the Adverse

Effects of Sharing the RWSTs of Both Reactor Units)

(2) Failure to Adequately Implement a Design Change Associated with the RWSTs

Introduction: The team identified a finding of very-low safety significance (Green),

and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control,

for the licensees failure to translate applicable design basis into TS Surveillance

Requirement (SR) 3.5.4.2 implementing procedures. Specifically, these procedures

did not verify RWST vent line was free of ice blockage at the locations and during all

applicable MODEs of reactor operation assumed by the ECCS and CS pump NPSH

calculation.

Description: Each reactor unit has one RWST, which supplies borated water to both

trains of the ECCS and CS systems during the injection phase of a LOCA recovery.

The TS 3.5.4, Refueling Water Storage Tank, required the RWSTs to be operable

when their associated reactor unit is in MODEs 1, 2, 3, or 4. A vent line is installed at

the top of each RWST. The vent lines are routed into the auxiliary building where they

connect to a common header which joins to a filtration system. Because the header is

common to both vents, the free air spaces of the RWSTs are communicated via their

vent lines. The vent line portions located between the tanks and the auxiliary building

are exposed to outside ambient conditions. For this reason, TS SR 3.5.4.2 stated,

Verify RWST vent path temperature is 35 degrees Fahrenheit. The associated TS

Basis explained that Heat traced portions of the RWST vent path should be verified to

be within the temperature limit needed to prevent ice blockage and subsequent vacuum

formation in the tank during rapid level decreases caused by accident conditions. The

licensee established procedures 1/2 BOSR 01-1,2,3, Modes 1, 2, and 3 Shiftily and

Daily Operating Surveillance, and 1/2 BOSR 01-4, Mode 4 Shiftily and Daily Operating

Surveillance, as the implementing procedures for SR 3.5.4.2.

Originally, the RWSTs design assumed they were atmospheric tanks by crediting their

associated vent line capability to prevent vacuum during tank drawdown. However, on

November 16, 2005, the licensee implemented a design change to credit the vent lines

capability to communicate the free air space of both tanks following the discovery that

the RWST vents did not have the capacity to prevent vacuum during the high outflow

expected during the injection phase, and the vent vacuum relief devices were not safety

related. This discovery was captured in the CAP as AR00239280.

As a result, calculation BYR 04-016, RHR, SI, CV and CS Pump NPSH during ECCS

Injection Mode, credited the vent lines of both RWSTs to mitigate the vacuum expected

during the drawdown of one tank during accident conditions. However, the team noted

this change was not correctly implemented into procedures 1/2 BOSR 01-1,2,3 and

1/2 BOSR 01-4. Specifically, these procedures were reactor unit specific in that their

instructions only required verifying the RWST vent line portions that were associated

20

with the applicable reactor unit RWST; that is, the portions between the associated

RWST and the auxiliary building. As a consequence, the team was concerned because,

if one vent line is found to be blocked with ice, the procedures would only recognize one

RWST as being inoperable. In addition, the procedures were only implemented when

the associated reactor unit was in MODEs 1, 2, 3, or 4 consistent with the applicability

requirements of TS 3.5.4. Thus, the team was also concerned that a potentially

inoperable condition would not be detected because the procedures would not verify

both vent lines were free of ice blockage when one reaction unit is in MODE 5 or 6

while the other reactor unit is in MODE 1, 2, 3, or 4.

The licensee captured the team concerns in their CAP as AR 02496766. The immediate

corrective action was to verify that outside air temperatures were not forecasted to fall

below 35 degrees Fahrenheit for the foreseeable future. Additionally, the licensee

determined the RWSTs remained operable during the last 3 years by performing a

historical review which did not find instances in which the vent lines temperature fell

below 35 degrees Fahrenheit. The proposed corrective actions to restore compliance

at the time of this inspection included revising the applicable calculations to remove

dependence on the opposite unit, and/or revising the affected procedures to be

consistent with the applicable calculation.

The team also noted the licensee did not perform a written safety evaluation that

provided the bases for the determination that this change, which resulted in a reduction

of reactor unit independence, did not require a license amendment. This separate

concern is discussed in detail in Section 1R21.5.b(1) of this report.

Analysis: The team determined the failure to translate applicable design basis into

TS SR 3.5.4.2 implementing procedures was contrary to 10 CFR Part 50, Appendix B,

Criterion III, Design Control, and was a performance deficiency. The performance

deficiency was determined to be more than minor because it was associated with the

Mitigating Systems cornerstone attribute of design control, and affected the cornerstone

objective of ensuring the availability, reliability, and capability of mitigating systems to

respond to initiating events to prevent undesirable consequences. Additionally, it was

associated with the Barrier Integrity cornerstone attribute of design control, and affected

the cornerstone objective of providing reasonable assurance that physical design

barriers protect the public from radionuclide releases caused by accidents or events.

Specifically, TS SR 3.5.4.2 implementing procedures were inadequate to verify RWST

operability because they did not verify all critical assumptions made by the design

calculations. The RWST supports ECCS, which is a mitigating system, and CS, which

is part of the physical design barrier.

The team determined the finding could be evaluated using the SDP in accordance

with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial

Characterization of Findings. Since the finding impacted the Mitigating Systems and

Barrier Integrity cornerstones, the inspectors screened the finding through IMC 0609,

Appendix A, The Significance Determination Process for Findings At-Power, using

Exhibit 2, Mitigating Systems Screening Questions, and Exhibit 3, Barrier Integrity

Screening Questions. The finding screened as very-low safety significance (Green)

because it did not result in the loss of operability or functionality, and it did not represent

an actual open pathway in the physical integrity of reactor containment. Specifically, the

licensee performed a historical review of the last 3 years of operation and did not find

any instances in which the vent path temperature fell below 35 degrees Fahrenheit.

21

The inspectors did not identify a cross-cutting aspect associated with this finding

because it was confirmed not to be reflective of current performance due to the age

of the performance deficiency. Specifically, the finding occurred approximately

10-years ago.

Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires,

in part, that design changes, including field changes, be subjected to design control

measures commensurate with those applied to the original design.

Contrary to the above, on November 16, 2005, the licensee performed a design change

and failed to subject it to design control measures commensurate to those applied to the

original design. Specifically, the licensee changed the ECCS and CS pump NPSH

calculation for their injection mode of operation (i.e., calculation BYR 04-016) to credit

the capability of the vent lines of both RWSTs to support the operability of any one

RWST. However, the design control measures failed to correctly translate the new

design basis into procedures 1/2 BOSR 01-1,2,3 and 1/2 BOSR 01-4 in that they were

not revised to verify the capability of the vent lines of both RWSTs to support the

operability of any one RWST.

The licensee is still evaluating its planned corrective actions. However, the team

determined that the continued non-compliance does not present an immediate safety

concern because outside air temperatures were not forecasted to fall below 35 degrees

Fahrenheit for the foreseeable future. Additionally, a corrective action tracking item was

created to develop compensatory actions if compliance is not restored prior to the next

season when temperatures can potentially decrease below 35 degrees Fahrenheit.

Because this violation was of very-low safety significance and was entered into the

licensees CAP as AR 02496766, this violation is being treated as a NCV, consistent

with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000454/2015008-05; 05000455/2015008-05; Failure to Adequately Implement a Design Change Associated

with the RWSTs)

(3) Failure to Evaluate the Adverse Effects of Changing the Emergency Service Water

Cooling Tower Tornado Analysis as Described in the Updated Final Safety Analysis

Report

Introduction: The team identified a Severity Level IV NCV of 10 CFR 50.59(d)(1),

Changes, Tests, and Experiments, and an associated finding of very-low safety

significance (Green) for the licensees failure to perform a written evaluation that

provided the bases for the determination that the changes to the SXCT tornado analysis

as described in the UFSAR did not require a license amendment. Specifically,

50.59 Evaluation 6G-11-0041, Tornado Missile Design Basis for the Essential Service

Water Cooling Towers, did not address the introduction of a new failure mode, the

resulting loss of heat removal capacity during worst postulated conditions, and addition

of operator actions that have not been demonstrated can be completed within the

required time to restore the required SXCT heat removal capacity during worst case

conditions.

Description: During the 2005 NRC Safety Systems Design, Performance and

Capability (SSDPC) inspection, the inspectors noted that the UFSAR-described

tornado analysis for the SXCT had not been updated to reflect changes that increased

the heat load. The SSDPC documented this concern as URI 05000454/2005002-07;

22 05000455/2005002-07. In 2007, this URI was subsequently closed to NCV 05000454/

2007004-03;05000455/2007004-03. As a result, on February 14, 2012, the licensee

completed EC 385829, UHS Capability with Loss of SX Fans Due to Tornado Missiles,

to change the UHS tornado missile design basis as described in Revision 7 of the

UFSAR. The EC 385829 evaluated these design basis changes in 10 CFR 50.59 safety

evaluation 6G-11-004, Tornado Missile Design Basis for the Essential Service Water

Cooling Towers, dated February 9, 2012. This 10 CFR 50.59 safety evaluation

concluded that the design basis changes could be implemented without obtaining a

license amendment.

However, the team noted that the licensee did not address the adverse effects of the

changes in the 10 CFR 50.59 safety evaluation. Specifically, the change reduced the

amount of missiles from multiple to single, and changed the SXCT design from

natural draft cooling to mechanical draft cooling (i.e., from passive to active system).

These changes adversely impacted 10 CFR 50.59 change evaluation criteria because

they would result in more than a minimal increase in the likelihood of occurrence of a

malfunction of the SXCT during a tornado event. Specifically:

The change introduced a new failure mode (i.e., fan failures) that was not

bounded by the previous analysis. Specifically, Revision 7 of the UFSAR

Section 3.5.4, Analysis of Multiple Missiles Generated by a Tornado, stated

that the SXCT fans, fan motors, and fan drives were not protected from tornado

missiles. It also stated, An analysis of cooling tower capacity without fans

[emphasis added] has been made. In contrast, this statement was revised to,

An analysis of the UHS cooling capability for a tornado missile event has been

made. The new analysis required multiple operating fans to ensure enough

cooling capacity to mitigate the effects of a single tornado missile. The fans, fan

motors, and fan drives were not modified to add tornado missile protection. In

addition, Revision 7 of the UFSAR Section 9.2.5.3.2, Essential Service Water

Cooling Towers, stated An analysis of the effect of multiple [emphasis added]

tornado missiles on the essential service water cooling towers has been

performed. This statement was revised to delete the word multiple.

Following this revision, the analysis only considered the effects of one

tornado-generated missile.

Revision 1 of NEI 96-07, Guidelines for 10 CFR 50.59 Evaluations, which

has been endorsed by the NRC in Regulatory Guide 1.187, Guidance for

Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, stated, in

part, that a change would result in less than a minimal increase in the likelihood

of occurrence of an SSC malfunction provided it satisfies applicable design

basis requirements. In contrast, this change did not satisfy the design basis

requirements for protection against natural phenomena as described in the

USAR Section 3.1.2.1.2, Evaluation Against Criterion 2 - Design Bases for

Protection Against Natural Phenomena. Specifically, Revision 7 and the

revision in effect at the time of this inspection of UFSAR Section 3.1.2.1.2 stated,

The systems, components, and structures important to safety have been

designed to accommodate, without loss of capability [emphasis added], effects of

the design-basis natural phenomena along with appropriate combinations of

normal and accident conditions. However, this change would result in the loss

of SXCT capability to perform its safety function during the worst case conditions

in that the required number of fans would not be available necessitating operator

23

actions to delay shutdown cooling initiation until an adequate number of SXCT

fans are available to support the shutdown cooling heat load and, consequently,

transition to MODE 5 where design basis accidents (DBAs) are not postulated.

The change involved a new operator action that supports the SXCT function

which is not reflected in plant procedures and training programs. Specifically,

UFSAR Section 3.5.4 was revised to credit new operator actions to delay

RHR initiation until an adequate number of SXCT fans are available for shutdown

cooling [emphasis added] and to stagger RHR initiation for the two units.

The revised UFSAR-described analysis assumed For the worst case design

conditions the first unit is assumed to be placed on RHR cooling 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after

the event and the second unit at 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after the event. NEI 96-07 states, in

part, that a new operator action that supports a design function credited in a

safety analysis results in less than a minimal increase in the likelihood of

occurrence of an SSC malfunction provided the action is reflected in plant

procedures and training programs, and these actions have been demonstrated

can be completed in the time required considering the aggregate effects.

However, the licensee had not created procedures and training material to

restore an adequate number of SXCT fans. In addition, the licensee had not

demonstrated that these actions can be completed in the time required

considering the aggregate effects, such as the expected conditions when the

actions are required.

In addition, the change would create a possibility for an SXCT malfunction with a

different result than any previously evaluated in the UFSAR because:

Nuclear Energy Institute (NEI) 96-07 states, A malfunction that involves an

initiator or failure whose effects are not bounded by those explicitly described in

the UFSAR is a malfunction with a different result. In contrast, this change

would result in the loss of SXCT capability to perform its safety function during

the worst case conditions in that the required number of fans would not be

available to support RHR initiation necessitating a delay of RHR initiation until an

adequate number of fans are available. The previous UFSAR-described analysis

assumed the SXCT design remained capable of performing its safety function

during the worst case conditions because it did not require any fans to support

RHR initiation and operation; and

NEI 96-07 stated, An example of a change that would create the possibility for a

malfunction with a different result is a substantial modification that creates a

new or common cause failure that is not bounded by previous analyses or

evaluations. In contrast, this change introduced a new failure that was not

bounded by previous analysis as previously explained.

The licensee captured the team concern in their CAP as AR 2506214 to request a

license amendment. The potential operability implications of this issue are discussed

in Section 4OA2.1.b(3) of this report.

Analysis: The team determined that the failure to perform a written evaluation that

provided the bases for the determination that the changes to the SXCT tornado analysis

as described in the UFSAR did not require a license amendment was contrary to the

requirements of 10 CFR 50.59(d)(1) and was a performance deficiency. The

24

performance deficiency was more than minor because it was associated with the

Mitigating Systems cornerstone attribute of protection against external events, and

affected the cornerstone objective of ensuring the availability, reliability, and capability of

mitigating systems to respond to initiating events to prevent undesirable consequences.

Specifically, the change did not ensure the SXCT reliability and availability during and

following a tornado event because it introduced a new failure mode, and added reliance

on operator actions that have not been demonstrated can be completed in the required

time. The change also did not ensure the SXCT capability to perform its safety function

during the worst case conditions during and following a tornado event in that the

required number of fans would not be available necessitating timely operator action

to restore the required heat removal capability.

In addition, the associated violation was determined to be more than minor because the

team could not reasonably determine the changes would not have ultimately required

NRC prior approval.

Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process

instead of the SDP because they are considered to be violations that potentially impede

or impact the regulatory process. This violation is associated with a finding that has

been evaluated by the SDP, and communicated with an SDP color reflective of the

safety impact of the deficient licensee performance. The SDP, however, does not

specifically consider the regulatory process impact. Thus, although related to a common

regulatory concern, it is necessary to address the violation and finding using different

processes to correctly reflect both the regulatory importance of the violation and the

safety significance of the associated finding.

In this case, the team determined the finding could be evaluated using the SDP in

accordance with IMC 0609, Significance Determination Process, Attachment 0609.04,

Initial Characterization of Findings. Because the finding impacted the Mitigating

System cornerstone, the team screened the finding through IMC 0609, Appendix A, The

Significance Determination Process for Findings At-Power, using Exhibit 2, "Mitigating

Systems Screening Questions." In accordance with Exhibit 2, the team screened the

finding using Exhibit 4, External Events Screening Questions, because the finding

involved the degradation of equipment or function specifically designed to mitigate a

severe weather initiating event. The team conservatively screened the finding as

necessitating a detailed risk evaluation because the loss of UHS during a tornado event

would degrade one or more trains of a system that supports a risk-significant system or

function.

The Senior Reactor Analysts (SRAs) performed a bounding risk evaluation for the delta

core damage frequency (CDF) of tornado missile strike(s) causing a core damage

event at Byron due to damage to the SXCT fans:

The SRAs assumed that a tornado with wind speed exceeding 100 mph would

be required to generate damaging missiles;

The frequency of this tornado for Byron is approximately 1.13E-4/yr from the Risk

Assessment Standardization Project (RASP) website;

25

The tornado missiles were assumed to cause damage and fail an entire set of

SXCT fans in addition to a set of fans that were initially out of service (i.e., 4 fans

conservative assumption); and

The SRAs further assumed that the tornado also caused a severe weather loss

of offsite power event.

The Byron SPAR Model Version 8.27 and Systems Analysis Programs for Hands-on

Integrated Reliability Evaluations (SAPHIRE) Version 8.1.2 software were used by the

SRAs to evaluate the risk significance of this finding. Using the Byron SPAR model, the

Conditional Core Damage Probability (CCDP) (i.e., if the tornado event occurred and

damaged one train of SXCT fans) is approximately 4.8E-4. Thus, a bounding CDF

calculated due to the SXCT vulnerability to missiles is approximately 5.4E-8/yr

(i.e., 1.13E-4/yr x 4.8E-4 = 5.4E-8/yr).

Based on the detailed risk evaluation, the SRAs determined that the finding was of

very-low safety significance (Green). As a result, this violation is categorized as

Severity Level IV in accordance with Section 6.1.d of the NRC Enforcement Policy.

The team did not identify a cross-cutting aspect associated with this finding because the

finding was not representative of current performance. Specifically, the change was

evaluated through the licensee 50.59 process in February 9, 2012.

Enforcement: Title 10 CFR 50.59, Changes, Tests, and Experiments, Section (d)(1)

requires, in part, the licensee to maintain records of changes in the facility, of changes in

procedures, and of tests and experiments made pursuant 10 CFR 50.59(c). These

records must include a written evaluation which provides the bases for the determination

that the change, test, or experiment does not require a license amendment pursuant to

paragraph (c)(2) of this section. Paragraph (c)(2)(ii) states, in part, that a licensee shall

obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed

change, test, or experiment if the change, test, or experiment would result in more than a

minimal increase in the likelihood of occurrence of a malfunction of an SSC important to

safety previously evaluated in the UFSAR. In addition, 10 CFR(c)(2)(vi) states, in part,

that a licensee shall obtain a license amendment pursuant to 10 CFR 50.90 prior to

implementing a proposed change, test, or experiment if the change, test, or experiment

would create a possibility for a malfunction of an SSC important to safety with a different

result than any previously evaluated in the Final Safety Analysis Report (FSAR)as

updated.

The UFSAR Section 9.2.5.3.2 in effect prior to the change implementation stated, An

analysis of the effect of multiple [emphasis added] tornado missiles on the essential

service water cooling towers has been performed. In addition, UFSAR Sections 3.5.4.1

and 9.2.5.3.2 in effect prior to the change implementation stated, An analysis of cooling

tower capacity without fans [emphasis added] has been made. Moreover, UFSAR

Section 3.1.2.1.2 in effect prior to the change implementation and at the time of this

inspection stated, The systems, components, and structures important to safety have

been designed to accommodate, without loss of capability [emphasis added], effects of

the design-basis natural phenomena along with appropriate combinations of normal and

accident conditions.

26

Contrary to the above, on February 9, 2012, the licensee failed to maintain a record of

a change in the facility made pursuant to 10 CFR 50.59(c) that included a written

evaluation which provided the bases for the determination that the change did not

require a license amendment pursuant to 10 CFR 50.59(c)(2). Specifically, the licensee

made changes to the UFSAR-described SXCT tornado analysis and evaluated this

change in 50.59 Evaluation 6G-11-0041. However, this evaluation did not consider

the adverse effects of the introduction of a new failure mode, the resulting loss of heat

removal capacity during worst postulated conditions, and addition of operator actions

that have not been demonstrated can be completed in the required time to restore the

required SXCT heat removal capacity during worst case conditions. As a result, the

evaluation did not provide a basis for the determination that the change did not result in

a more than a minimal increase in the likelihood of occurrence of a malfunction of the

SXCT during and following a tornado event, and would not create a possibility for a

malfunction of the SXCT with a different result than any previously evaluated.

The licensee is still evaluating its planned corrective actions to restore compliance. As

an immediate corrective action, the licensee performed an operability evaluation. At the

time of the CDBI exit meeting on June 16, 2015, the team was still reviewing the revised

operability evaluation with the assistance of NRR.

Because this was a Severity Level IV violation, and was entered into the licensees CAP

as AR 02506214, this violation is being treated as an NCV, consistent with Section 2.3.2

of the NRC Enforcement Policy. (NCV 05000454/2015008-06; 05000455/2015008-06,

Failure to Evaluate the Adverse Effects of Changing the SXCT Tornado Analysis as

Described in the UFSAR)

The associated finding is evaluated separately from the traditional enforcement violation

and, therefore, the finding is being assigned a separate tracking number.

(FIN 05000454/2015008-07; 05000455/2015008-07, Failure to Evaluate the Adverse

Effects of Changing the SXCT Tornado Analysis as Described in the UFSAR)

.6

Operating Procedure Accident Scenarios

a.

Inspection Scope

The team performed a detailed reviewed of the procedures listed below. The

procedures were chosen because they were associated with feed-and-bleed of the

RCS, a loss of UHS, and other aspects of this inspection. For the procedures listed

time critical operator actions were reviewed for reasonableness, in plant action were

walked down with a licensed operator, and any interfaces with other departments were

evaluated. The procedures were compared to the UFSAR, design assumptions, and

training materials to assess consistency.

The following operating procedures were reviewed in detail:

1BFR-H1, Response to Loss of Secondary Heat Sink Unit1, Revision 203;

0BOA PRI-7, Loss of Ultimate Heat Sink Unit 0, Revision 1;

1BOA PRI-7, Essential Service Water Malfunction Unit 1, Revision 106;

1BOA PRI-5, Control Room Inaccessibility, Revision 108;

27

1BOA ELEC-5, Local Emergency Control of Safe Shutdown Equipment,

Revision 106;

1BEP ES-1.3, Transfer to Cold Leg Recirculation Unit 1, Revision 204; and

1BCA-1.2, LOCA Outside Containment Unit 1, Revision 200.

b.

Findings

(1) Failure to Provide Proper Direction for Low Level Isolation of the Refueling Water

Storage Tank in Emergency Operating Procedures

Introduction: The team identified a finding of very-low safety significance (Green), and

an associated NCV of TS 5.4, Procedures, for the failure to EOPs for transfer to cold

leg recirculation. Specifically, Revision 204 of EOPs 1/2BEP ES-1.3, Transfer to Cold

Leg Recirculation, did not contain instructions for transferring the ECCS and CS

systems to the recirculation mode that ensured prevention of potential pump damage

when the RWST is emptied following a LOCA.

Description: Procedures 1/2BEP ES-1.3 were established as the implementing EOPs

for transferring ECCS and CS system suction from the RWST to containment sump

recirculation. These EOPs were intended to be consistent with the technical guidelines

of Westinghouse Owners Group Guidelines (WOG) Emergency Response Guideline

(ERG) ES-1.3, Transfer to Cold Leg Recirculation, dated April 30, 2005. The technical

guideline of WOG ERG ES-1.3 included the following caution statement: Any pumps

taking suction from the RWST should be stopped if RWST level decreases to (U.03).

The ERG defined (U.03) as RWST empty alarm set point in plant specific units. It also

stated, Based on pump suction piping configuration, the plant specific value of (U.03)

may need to consider the possibility of vortexing and air entrainment. The ERG basis

for this caution stated, Any pumps taking suction from the RWST must be stopped

when the level in the tank reaches the empty alarm set point in order to prevent loss of

suction flow and potential pump damage. The licensee established 9 percent RWST

level as the empty alarm set point to prevent air-entraining vortices and ensured

adequate pump NPSH.

In 1996, the licensee changed EOPs 1/2BEP ES-1.3 to include a deviation to this ERG

caution. Specifically, the revised EOP caution stated Any pumps taking suction from

the RWST should be stropped if level drops to 9 percent, unless a flow path also exists

from the CNMT [containment] sump. The EOP deviation document stated This will

allow continuing with switchover without securing pumps if an acceptable flow path

exists. It also stated CNMT pressure should isolate the RWST flow path once aligned

to the sump. However, the licensee did not perform any evaluation to support this

rationale.

The team was concerned because the revised caution did not assure to prevent air

entrainment into the piping system to avoid ECCS and CS pump air binding and/or

cavitation leading to potential damage. The licensee captured the team concern in

their CAP as AR 02495580. The immediate corrective action was to create a standing

order instructing operators to secure all pumps aligned to the RWST when it reaches

9 percent level. The proposed corrective actions to restore compliance at the time of

this inspection included performing a detailed engineering analysis of the hydrodynamic

fluid mechanics with a dual suction source option or removing the dual suction source

option.

28

Analysis: The team determined that the failure to maintain an EOP for transfer to cold

leg recirculation was contrary to TS 5.4, Procedures, and was a performance

deficiency. The performance deficiency was determined to be more than minor because

it was associated with the Mitigating Systems cornerstone attribute of procedure quality,

and affected the cornerstone objective of ensuring the availability, reliability, and

capability of mitigating systems to respond to initiating events to prevent undesirable

consequences. In addition, it was associated with the Barrier Integrity cornerstone

attribute of procedure quality, and affected the cornerstone objective of providing

reasonable assurance that physical design barriers protect the public from radionuclide

releases caused by accidents or events. Specifically, failure to maintain an EOP for

transfer to cold leg recirculation does not ensure that air entrainment into the piping

system is prevented. As a consequence, the availability, reliability, and capability of

the ECCS pumps to meet their mitigating function are not ensured. Similarly, the

performance deficiency does not provide reasonable assurance the CS pumps would

remain capable of supporting the reactor containment barrier function.

The team determined the finding could be evaluated using the SDP in accordance

with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial

Characterization of Findings. Because the finding impacted the Mitigating Systems

and Barrier Integrity cornerstones, the team screened the finding through IMC 0609,

Appendix A, The Significance Determination Process for Findings At-Power, using

Exhibit 2, Mitigating Systems Screening Questions, and Exhibit 3, Barrier Integrity

Screening Questions. The finding screened as of very-low safety significance (Green)

because it did not result in the loss of operability or functionality of mitigating systems,

represent an actual open pathway in the physical integrity of reactor containment, and

involved an actual reduction in function of hydrogen igniters in the reactor containment.

Specifically, the incorrect caution would only be used in the event that transfer to sump

recirculation was not completed by 9 percent tank level or if the RWST suction isolation

valves fail to close. With respect to transfer to sump recirculation by 9 percent tank

level, this is a time critical operator action that is tested and verified periodically on the

plant simulator. A review of these simulator test results reasonably determined that

operators reliably complete the transfer to sump recirculation prior to reaching this set

point. With respect to the failure of the RWST suction isolation valves, these valves are

test quarterly to demonstrate operability. A review of these test results for the last

3 years reasonably determined the valves would have isolated the tank when required.

The team did not identify a cross-cutting aspect associated with this finding because the

finding was not representative of current performance. Specifically, the inadequate

caution had been added to 1/2BEP ES-1.3 in 1996.

Enforcement: In TS Section 5.4.1b states, in part, that written procedures shall be

established, implemented, and maintained covering the EOPs required to implement the

requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic

Letter (GL) 82-33, Section 7.1. NUREG-0737, Supplement 1, Section 7.1.c, states,

Upgrade EOPs to be consistent with Technical Guidelines and an appropriate

procedure Writers Guide. The applicable technical guideline contained in WOG ERG

ES-1.3, Transfer to Cold Leg Recirculation, dated April 30, 2005, stated, Any pumps

taking suction from the RWST should be stopped if RWST level decreases to (U.03).

The ERG defined (U.03) as RWST empty alarm set point in plant specific units. It also

stated, Based on pump suction piping configuration, the plant specific value of (U.03)

may need to consider the possibility of vortexing and air entrainment.

29

The licensee established Revision 204 of 1/2BEP ES-1.3, Transfer to Cold Leg

Recirculation, as the implementing procedures for WOG ERG ES-1.3 to specify the

actions required for transfer to containment sump recirculation. In addition, the licensee

established 9 percent RWST level as the empty alarm set point, in part, to prevent air

entrainment.

Contrary to the above, between 1996 to at least May 4, 2015, the licensee failed to

maintain a written procedure covering the EOPs required to implement the requirements

of NUREG-0737 and NUREG-0737, Supplement 1, as stated in GL 82-33, Section 7.1.

Specifically, the licensee did not upgrade EOPs 1/2BEP ES-1.3 to be consistent with the

technical guideline contained in WOG ERG ES-1.3 in that the EOPs did not instructed

operators to stop any pumps taking suction from the RWST if level decreases below the

9 percent RWST empty alarm set point when a flow path from the containment sump

existed.

The licensee is still evaluating its planned corrective actions. However, the team

determined that the continued non-compliance does not present an immediate safety

concern because the licensee created a standing order instructing operators to secure

all pumps aligned to the RWST when it reaches 9 percent level.

Because this violation was of very-low safety significance, and was entered into the

licensees CAP as AR 02495580, this violation is being treated as an NCV, consistent

with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000454/2015008-08; 05000455/2015008-08, Failure to Provide Proper Direction for Low Level Isolation of

the RWST in EOPs)

4.

OTHER ACTIVITIES

4OA2 Identification and Resolution of Problems

.1

Review of Items Entered Into the Corrective Action Program

a.

Inspection Scope

The team reviewed a sample of the selected component problems identified by the

licensee, and entered into the CAP. The team reviewed these issues to verify an

appropriate threshold for identifying issues, and to evaluate the effectiveness of

corrective actions related to design issues. In addition, corrective action documents

written on issues identified during the inspection were reviewed to verify adequate

problem identification and incorporation of the problem into the CAP. The specific

corrective action documents sampled and reviewed by the team are listed in the

attachment to this report.

The team also selected three issues identified during previous CDBIs to verify that

the concern was adequately evaluated and corrective actions were identified and

implemented to resolve the concern, as necessary. The following issues were reviewed:

NCV 05000454/2012007-01; 05000455/2012007-01, Non-Conforming 480/120

VAC Motor Control Contactors;

NCV 05000454/2012007-03; 05000455/2012007-03, Non-Conservative

Calibration Tolerance Limits for Electrical Relay Settings; and

30

NCV 05000454/2012007-05; 05000455/2012007-05, Failure to Provide Means

to Detect Leak in Emergency Core Cooling Flow Path.

b.

Findings

(1) Failure to Promptly Correct an NRC-Identified Non-Cited Violation Associated with the

Capability to Detect and Isolate Emergency Core Cooling System Leakage

Introduction: A finding of very-low safety significance (Green), and an associated cited

violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, was

identified by the team for the failure to correct a condition adverse to quality (CAQ).

Specifically, on June 15, 2012, the NRC issued an NCV for the failure to provide means

to detect and isolate a leak in the ECCS within 30 minutes as described in the UFSAR,

which is a CAQ. As of May 22, 2015, the licensee had not corrected the CAQ.

Description: On June 15, 2012 the NRC identified that the licensee had failed to provide

a means to detect and isolate a leak in the ECCS flow path within 30 minutes, as

described in UFSAR 6.3.2.5, System Reliability. Specifically, UFSAR 6.3.2.5 stated,

in part, that the design of the auxiliary building and related equipment was based upon

handling of leaks up to a maximum of 50 gallons per minute (gpm). In addition, it stated

Means were provided to detect and isolate such leaks in the emergency core cooling

flow path within 30 minutes. The 2012 CDBI team identified that the licensee had failed

to provide a means to detect and isolate an ECCS leak within 30 minutes. This issue

was documented as NCV 05000454/2012007-05; 05000455/2012007-05, Failure to

Provide Means to Detect Leak in ECCS Flow Path, in Inspection Report (IR) 05000454/

2012007; 05000455/2012007.

The licensee captured this NCV in their CAP as AR 01378257 and AR 01398434. The

assigned corrective action tracking item (CA) was AR01378257-04, which stated:

Investigate the bases/sources of the values assigned to the single failure

(50 gpm and 30 minutes), including whether there is a commitment associated.

Create additional corrective actions (CA type) as necessary. If UFSAR change is

determined feasible, include an action to determination of the impact of the leak

duration lasting longer than 30 minutes on flood level inside containment and the

Auxiliary Building.

The CA due date was extended eight times and, eventually, the CA was downgraded to

an action tracking item (ACIT) because the licensee recognized that it did not correct the

issue. Procedure PI-AA-125, Corrective Action Program Procedure, defined ACIT as

Action items that are completed to improve performance, or correct minor problems that

do not represent CAQ. On February 18, 2015, the licensee discovered that a new CA

type assignment was not generated to address the NCV following the AR 01378257-04

downgrade from a CA to an ACIT type. This was inconsistent with step 4.5.2 of

procedure PI-AA-125 in that it required, in part, the creation of a CA for any planned

action necessary to correct a CAQ. This discovery was captured in the CAP as

AR 02454767. The associated CA assignment stated:

Design Engineering will determine if UFSAR section 6.3.2.5 requires revision

using the information provided in IR 01378257 and IR 1398434. If it is concluded

a revision is required, an additional CA to track the change will be created.

31

During this inspection period, the team noted that the actions assigned by this CA were

similar to those of AR 01378257-04, which the licensee had previously determined

did not correct the NCV. The team was concerned because, as of May 22, 2015, the

licensee failed to restore compliance and failed to have objective plans to restore

compliance in a reasonable period following the NRC identification of the NCV on

June 15, 2012.

The licensee captured the teams concern in their CAP as AR 02501454 to promptly

restore compliance. As an immediate corrective action, the licensee reasonably

determined ECCS remained operable by reviewing procedures and calculations.

Specifically, the licensee reasonably determined procedures used when responding to

postulated events would direct operators to detect and isolate an ECCS leak before it

could adversely affect the system mitigating function or result in a radionuclide release

in excess of applicable limits.

Analysis: The team determined that the failure to correct an NRC-identified NCV

associated with the capability to detect and isolate ECCS leakage, which is a CAQ, was

contrary to 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, and was a

performance deficiency. The performance deficiency was determined to be more than

minor because it was associated with the Mitigating Systems cornerstone attribute of

design control, and affected the cornerstone objective of ensuring the availability,

reliability, and capability of mitigating systems to respond to initiating events to prevent

undesirable consequences. In addition, it was associated with the Barrier Integrity

cornerstone attribute of design control, and affected the cornerstone objective of

providing reasonable assurance that physical design barriers protect the public from

radionuclide releases caused by accidents or events. Specifically, the failure to detect

and isolate a leak in the ECCS flow path within 30 minutes could compromise long term

cooling, adversely affecting its capability to mitigate a DBA. In addition, a detection and

isolation time greater than the time assumed by the design basis for an ECCS leak

following an accident would result in greater radionuclide release to the auxiliary

building, and the environment and, thus, does not assure that physical design barriers

protect the public from radionuclide releases caused by accidents or events.

The team determined the finding could be evaluated using the SDP in accordance

with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial

Characterization of Findings. Because the finding impacted the Mitigating Systems

and Barrier Integrity cornerstones, the team screened the finding through IMC 0609,

Appendix A, The Significance Determination Process for Findings At-Power, using

Exhibit 2, Mitigating Systems Screening Questions, and Exhibit 3, Barrier Integrity

Screening Questions. The finding screened as very-low safety significance (Green)

because it did not result in the loss of operability or functionality, and it did not represent

an actual pathway in the physical integrity of reactor containment. Specifically, the

licensee reasonably demonstrated that an ECCS leak could be detected and isolated

before it could adversely affect long-term cooling of the plant.

The team determined that the associated finding had a cross-cutting aspect in the area

of human performance because the licensee did not use a consistent and systematic

approach to make decisions. Specifically, the licensee downgraded the original CA to

an ACIT without creating a new CA, which was inconsistent with the instructions

contained in procedure PI-AA-125. Additionally, when the licensee subsequently

discovered a CA type assignment was not created to address the NCV, the licensee

32

created a CA assignment to track actions that were similar to those tracked by the ACIT,

which was inconsistent with the licensee previous determination that those actions did

not correct the NCV. [H.13]

Enforcement: Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action,

states, in part, that measures shall be established to assure that conditions adverse to

quality, such as failures, malfunctions, deficiencies, deviations, defective material and

equipment, and non-conformances are promptly identified and corrected.

Contrary to the above, from June 15, 2012, to at least May 22, 2015, the licensee failed

to correct a CAQ. Specifically, on June 15, 2012, the NRC issued NCV 05000454

/2012007-05;05000455/2012007-05 for the failure to provide means to detect and

isolate a leak in the ECCS within 30 minutes for Byron Station, Units 1 and 2, as

described in UFSAR Section 6.3.2.5, which is a CAQ. As of May 22, 2015, the

licensee had not corrected the CAQ in a reasonable period. Instead, the licensee

created ACTI to develop a plan to correct the CAQ, and the associated due date was

extended at least eight times.

The licensee is still evaluating corrective actions. However, the team determined that

the continued non-compliance does not present an immediate safety concern because

the licensee reasonably demonstrated that a leak could be detected and isolated before

it could adversely affect long-term cooling of the plant or result in a radionuclide release

in excess of applicable limits.

This violation is being cited as described in the Notice, which is enclosed with this IR.

This is consistent with the NRC Enforcement Policy, Section 2.3.2.a.2, which states, in

part, that the licensee must restore compliance within a reasonable period of time (i.e., in

a timeframe commensurate with the significance of the violation) after a violation is

identified. The NRC identified NCV 05000454/2012007-05; 05000455/2012007-05 on

June 15, 2012, and documented it in IR 05000454/2012007. The team determined that

the licensee failed to restore compliance within a reasonable time following issuance of

this NCV and failed to have objective plans to restore compliance. (VIO 05000454

/2015008-09;05000455/2015008-09, Failure to Promptly Correct an NRC-Identified

NCV Associated with the Capability to Detect and Isolate ECCS Leakage)

(2) Failure to Maintain the Accuracy of the Instrument Loops Used to Verify Compliance with

the Containment Average Air Temperature Technical Specification Limit

Introduction: The team identified a finding of very-low safety significance (Green), and

an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions,

Procedures, and Drawings, for the failure to have procedures to maintain the accuracy

within the necessary limits of instrument loops used to verify compliance with the

containment average air temperature TS limit of 120 degrees Farhenheit. Specifically,

in 2007, the licensee cancelled the periodic PMs intended to maintain the instrument

accuracy necessary for verifying compliance with the limiting condition for operation

(LCO) of TS 3.6.5, Containment Air Temperature.

Description: The team reviewed selected corrective action documents initiated by the

licensee as a result of their recent Focused Self-Assessment titled, Readiness Review

for 2015 NRC Component Design Basis Inspection. The reviewed corrective action

document sample included AR 02437973. This corrective action document was initiated

on January 15, 2015, in part, for the discovery that the four instrument loops used for

33

determining containment average air temperature (i.e., loops 1/2VP-030, 1/2VP-031,

1/2VP-032, and 1/2VP-033) were removed from the PM Program in 2007 via Service

Request 47654. The corrective action document also noted that the PMs were last

performed in 2001 for 1VP-030; 2002 for 1/2VP-031, 1/2VP-032, 2VP-030, and

2VP-033; and 2009 for 1VP-033.

This corrective action document created an ACIT to determine if the PMs should be

reestablished. Procedure PI-AA-125, Corrective Action Program Procedure, defined

ACIT as Action items that are completed to improve performance, or correct minor

problems that do not represent CAQ. On March 3, 2015, the ACIT concluded that there

was no need to reestablish the PMs due to the instrument loop reliability, previous

calibration history, loop design, redundancy, and daily monitoring which the licensee

believed would notice instrument drift. However, the team noted that TS SR 3.6.5.1

required verifying containment air temperature is less than 120 degrees Fahrenheit

by averaging the instrument readings and, thus, instrument reading variability was

expected. In addition, the team noted the licensee had not established a variability limit

(i.e., acceptance criteria) among the instrument loops and relied on operator judgment to

identify adverse drifts.

The team was concerned because these instrument loops were not maintained to

ensure their accuracy was within the necessary limits to verify compliance with the

containment average air temperature TS limit of 120 degrees Fahrenheit. Containment

average air temperature is an initial condition used in DBA analyses, and is an important

consideration in establishing the containment environmental qualification operating

envelope for both pressure and temperature. This TS limit ensures that initial conditions

assumed in these analyses are met during unit operations.

The licensee captured the teams concern in their CAP as AR 02502846. As an

immediate corrective action, the licensee reasonably established that the 120 degrees

Fahrenheit limit was not exceeded by reviewing applicable historical records from 2002

to time of this inspection. The proposed corrective action to restore compliance at the

time of this inspection was to reconstitute PM procedures for these instrument loops to

assure they are maintained.

Analysis: The team determined that the failure to have procedures to maintain the

accuracy within necessary limits of the instrument loops used during SR 3.6.5.1 was

contrary to 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and

Drawings, and was a performance deficiency. The performance deficiency was

determined to be more than minor because it was associated with the configuration

control attribute of the Barrier Integrity Cornerstone, and adversely affected the

cornerstone objective to ensure that physical design barriers protect the public from

radionuclide releases caused by accidents or events. Specifically, the failure to have

procedures to maintain the accuracy of the containment air temperature instrumentation

loops within necessary limits does not ensure the instrument loop accuracy is

maintained such that SR 3.6.5.1 activities are effective at verifying compliance with the

containment average air temperature TS limit. As a result, the potential exists for an

inoperable condition to go undetected.

The team determined the finding could be evaluated using the SDP in accordance

with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial

Characterization of Findings. Because the finding impacted the Barrier Integrity

34

cornerstone, the team screened the finding through IMC 0609, Appendix A, The

Significance Determination Process for Findings At-Power, using Exhibit 3, Barrier

Integrity Screening Questions. The finding screened as of very-low safety significance

(Green) because it did not represent an actual open pathway in the physical integrity of

reactor containment or involved an actual reduction in hydrogen igniter function.

Specifically, the containment integrity remained intact and the finding did not impact

the hydrogen igniter function.

The team determined that this finding had a cross-cutting aspect in the area of problem

identification and resolution because the licensee did not identify issues completely and

accurately in accordance with the CAP. Specifically, on January 15, 2015, the licensee

captured the lack of periodic PM activities for the containment air temperature instrument

loops in the CAP. However, the licensee failed to completely and accurately identify the

issue in that it was not treated as a CAQ. As a consequence, no corrective actions were

implemented. [P.1]

Enforcement: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,

and Drawings, requires, in part, that activities affecting quality be prescribed by

documented procedures of a type appropriate to the circumstances and be

accomplished in accordance with these procedures.

Contrary to the above, since 2007 to at least May 22, 2015, the licensee failed to have a

procedure for maintaining the accuracy within the necessary limits of the instrument

loops used while implementing SR 3.6.5.1. Specifically, in 2007, the licensee cancelled

the PMs intended to maintain the instrument loops accuracy necessary for verifying

compliance with LCO 3.6.5 limit.

The licensee is still evaluating its planned corrective actions. However, the team

determined that the continued non-compliance does not present an immediate safety

concern because containment average air temperature readings were significantly lower

than the associated TS limit, and are reasonably expected to maintain that margin in the

foreseeable future based on past performance.

Because this violation was of very-low safety significance, and was entered into the

licensees CAP as AR 02502846, this violation is being treated as an NCV, consistent

with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000454/2015008-10; 05000455/2015008-10, Failure to Maintain the Instrument Loops Used to Verify

Compliance with the Containment Average Air Temperature TS Limit)

(3) Operability Evaluation Relied on Probabilities of Occurrence of the Associated Event

Introduction: The team identified a finding of very-low safety significance (Green),

and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions,

Procedures, and Drawings, for the failure to make an operability determination without

relying on the use of probabilistic tools. Specifically, an operability evaluation related to

an SXCT degraded condition used probabilities of occurrence of tornado events which

was contrary to the requirements of Revision 16 of procedure OP-AA-108-115,

Operability Determinations.

Description: Revision 7 of UFSAR Section 3.5.4, Analysis of Multiple Missiles

Generated by a Tornado, stated that the SXCT fans, fan motors, and fan drives were

not protected from tornado missiles. It also stated that An analysis of cooling tower

35

capacity without fans has been made. In addition, it stated that Using the most

conservative design conditions, it is predicted if the plant is shut down under non-LOCA

conditions with loss of offsite power, the temperature of the service water supplied to

the plant will not exceed 110 degrees Farhernheit. However, during the 2005 NRC

SSDPC inspection, the inspectors noted that this analysis had not been updated to

reflect changes that increased the heat load. The SSDPC documented this concern as

URI 05000454/2005002-07; 05000455/2005002-07. In 2007, this URI was subsequently

closed to NCV 05000454/2007004-03; 05000455/2007004-03. As a result, on February

14, 2012, the licensee completed EC 385829, UHS Capability with Loss of SX Fans

Due to Tornado Missiles, to change the UHS tornado missile design basis to require

a minimum of two SXCT fans and motors for cooling following a tornado event. The

change did not include adding tornado protection to the fans, fan motors, and fan drives.

On August 9, 2013, the licensee initiated corrective action document IR 01545153 for

the NRC discovery that the associated written safety evaluation intended to provide

the bases for the determination that this change did not require a license amendment

failed to consider the change adverse effects. On August 14, 2013, the licensee

initiated corrective action document AR 1546621 to address the associated technical

implications. This corrective action document resulted in Revision 0 of Operability

Evaluation 13-007, Ultimate Heat Sink Capability with Loss of Essential Service Water

Cooling Tower Fans, intended to reasonably demonstrate UHS operability until

corrective actions to restore compliance were implemented.

During this inspection period, the CDBI team noted that Operability Evaluation 13-007

relied on the probability of occurrence of a tornado. Specifically, it stated The UHS is

capable of providing the required cooling because, given a tornado strike under the

design conditions in the UFSAR, the probability of occurrence is less than the

acceptance criteria of 10E-7 /year in SRP 2.2.3. It also stated that The software used

to determine the missile hit probability is called [Tornado Missile Risk Evaluation

Methodology] TORMIS. In addition, it stated that The software uses site specific

factors such as predicted tornado characteristics, tornado occurrence rates, building

layout, potential missile sources and types, missile distribution and the number of

potential missiles. The supporting analysis used the UFSAR Section 2.3.1.2.2,

Tornadoes and Severe Winds. tornado probability of occurrence value of 21E-4 per

year.

Procedure OP-AA-108-115, Operability Determinations, Section 4.5.13, Use of PRA,

stated:

PRA is a valuable tool for evaluating accident scenarios because it can consider

the probabilities of occurrence of accidents or external events. Nevertheless, the

definition of operability is that the SSC must be capable of performing its

specified function or functions, which inherently assumes that the event occurs

and that the safety function or functions can be performed. Therefore, the use of

PRA or probabilities of occurrence of accidents or external events is not

consistent with the assumption that the event occurs, and is not acceptable for

making operability decisions.

Thus, the team determined that the use of TORMIS, the probability for occurrence of

tornados, and the probabilities of missile strikes was not acceptable and contrary to

licensee procedure OP-AA-108-115. The team, in consultation with NRR, also

36

determined that this procedure requirement was consistent with Attachment C.06 of

NRC IMC 0326, Operability Determinations & Functionality Assessments for Conditions

Adverse to Quality or Safety, which was established to assist NRC inspectors review of

licensee determinations of operability and resolution of degraded or nonconforming

conditions.

In addition, the team noted that Byron had not obtained NRC approval for the site

specific use of TORMIS as stated in Regulatory Issue Summary (RIS) 2008-14, Use of

TORMIS Computer Code for Assessment of Tornado Missile Protection. Specifically,

the RIS stated that The initial use of the TORMIS methodology as described in this

RIS requires a license amendment in accordance with 10 CFR 50.59(c)(2)(viii) and

subsequent revision to the plant licensing basis because it is a Departure from the

method of evaluation described in the FSAR, as updated, used in establishing the

design bases or in the safety analysis as defined in 10 CFR 50.59(a)(2).

The team was concerned because Operability Evaluation 13-007 did not reasonably

demonstrate the degraded UHS would be capable of performing its function following a

tornado event. The licensee captured the team concern in their CAP as AR 2504624 to

revise Operability Evaluation 13-007 without using PRA tools.

Analysis: The team determined that the failure to make an operability determination

without relying on the use of probabilistic tools was contrary to licensee procedure

OP-AA-108-115 and was a performance deficiency. The performance deficiency was

determined to be more than minor because it was associated with the Mitigating

Systems cornerstone attribute of protection against external events, and affected the

cornerstone objective of ensuring the availability, reliability, and capability of mitigating

systems to respond to initiating events to prevent undesirable consequences.

Specifically, failure to perform an adequate operability evaluation does not ensure the

SXCT would remain capable of performing its safety function, and had the potential to

allow an inoperable condition to go undetected.

The team determined the finding could be evaluated using the SDP in accordance

with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial

Characterization of Findings. Because the finding impacted the Mitigating System

cornerstone, the team screened the finding through IMC 0609, Appendix A, The

Significance Determination Process for Findings At-Power, using Exhibit 2, "Mitigating

Systems Screening Questions." In accordance with Exhibit 2, the team screened the

finding using Exhibit 4, External Events Screening Questions, because the finding

involved the degradation of equipment or function specifically designed to mitigate a

severe weather initiating event. The team conservatively screened the finding as

necessitating a detailed risk evaluation because the loss of UHS during a tornado event

would degrade one or more trains of a system that supports a risk-significant system or

function.

The SRAs performed a bounding risk evaluation for the CDF of tornado missile

strike(s) causing a core damage event at Byron due to damage to the SXCT fans:

The SRAs assumed that a tornado with wind speed exceeding 100 mph would

be required to generate damaging missiles.

The frequency of this tornado for Byron is approximately 1.13E-4/yr from the

RASP website;

37

The tornado missiles were assumed to cause damage and fail an entire set of

SXCT fans in addition to a set of fans that were initially out of service (i.e., 4 fans

- conservative assumption); and

The SRAs further assumed that the tornado also caused a severe weather loss

of offsite power event.

The Byron SPAR Model Version 8.27 and SAPHIRE Version 8.1.2 software were used

by the SRAs to evaluate the risk significance of this finding. Using the Byron SPAR

model, the CCDP (i.e., if the tornado event occurred and damaged one train of SXCT

fans) is approximately 4.8E-4. Thus, a bounding CDF calculated due to the SXCT

vulnerability to missiles is approximately 5.4E-8/yr (i.e., 1.13E-4/yr x 4.8E-4 = 5.4E-8/yr).

Based on the detailed risk evaluation, the SRAs determined that the finding was of very

low safety significance (Green).

The team determined that this finding had a cross-cutting aspect in the area of human

performance because the licensee did not ensure knowledge transfer to maintain a

knowledgeable and technically competent workforce. Specifically, the licensee did not

ensure personnel were trained on the prohibition of the use of probabilities of occurrence

of an event when performing operability evaluations, which was contained in procedure

OP-AA-108-115. [H.9]

Enforcement: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,

and Drawings, requires, in part, that activities affecting quality be prescribed by

documented procedures of a type appropriate to the circumstances and be

accomplished in accordance with these procedures.

The licensee established Revision 16 of procedure OP-AA-108-115, Operability

Determinations, as the implementing procedure for assessing operability of SSCs, an

activity affecting quality. Section 4.5.13, Use of Probabilistic Risk Assessment, stated

[] the use of PRA or probabilities of occurrence of accidents or external events is not

consistent with the assumption that the event occurs, and is not acceptable for making

operability decisions.

Contrary to the above, on August 20, 2013, the licensee failed to follow Section 4.5.13 of

procedure OP-AA-108-115. Specifically, the licensee used a PRA tool (i.e., TORMIS)

and probabilities of occurrence of an external event (i.e., tornado) when making an

operability decision related to the SXCT degradation when mitigating tornado events.

Establishing a reasonable expectation of operability is an activity affecting quality.

As an immediate corrective action, the licensee revised the affected operability

evaluation without using PRA tools. At the time of the CDBI exit meeting on

June 16, 2015, the team was still reviewing the revised operability evaluation with

the assistance of NRR.

Because this violation was of very-low safety significance and was entered into the

licensees CAP as AR 2504624, this violation is being treated as an NCV, consistent

with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000454/2015008-11; 05000455/2015008-11, Operability Evaluation Relied on Probabilities of Occurrence of

the Associated Event)

38

4OA6 Management Meetings

.1

Interim Exit Meeting Summary

On May 22, 2015, the team presented the inspection results to Mr. R. Kearney, and

other members of the licensee staff. The licensee acknowledged the issues presented.

The inspectors had outstanding questions that required additional review and a follow-up

exit meeting.

.2

Exit Meeting Summary

On June 16, 2015, the team presented the inspection results to Mr. B. Currier, and other

members of the licensee staff. The licensee acknowledged the issues presented. The

team asked the licensee whether any materials examined during the inspection should

be considered proprietary. Several documents reviewed by the team were considered

proprietary information and were either returned to the licensee or handled in

accordance with NRC policy on proprietary information.

ATTACHMENT: SUPPLEMENTAL INFORMATION

Attachment

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

R. Kearney, Site Vice President

T. Chalmers, Plant Manager

C. Keller, Engineering Director

B. Currier, Senior Manager of Design Engineering

D. Spitzer, Regulatory Assurance Manager

J. Cunzeman, Mechanical/Structural Design Manager

A. Corrigan, NRC Coordinator

U.S. Nuclear Regulatory Commission

C. Lipa, Chief, Engineering Branch 2

J. Ellegood, Chief, Reactor Projects Branch 3 (Acting)

N. Féliz Adorno, Senior Reactor Inspector

C. Zoia, Senior Resident Inspector (Acting)

J. Draper, Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened 05000454/2015008-01; 05000455/2015008-01

URI

Question Regarding the Maximum Wet Bulb

Temperature Value Assumed in the SXCT Tornado

Analysis (Section 1R21.3.b(1))05000454/2015008-02; 05000455/2015008-02

URI

Maximum Wet Bulb Temperature Value Assumed in

SXCT Analysis Was Not Monitored

(Section 1R21.3.b(2))05000454/2015008-03; 05000455/2015008-03

NCV

Failure to Evaluate the Adverse Effects of Sharing the

RWSTs of Both Reactor Units (Section 1R21.5.b(1))05000454/2015008-04; 05000455/2015008-04

FIN

Failure to Evaluate the Adverse Effects of Sharing the

RWSTs of Both Reactor Units (Section 1R21.5.b(1))05000454/2015008-05; 05000455/2015008-05

NCV

Failure to Adequately Implement a Design Change

Associated with the RWSTs (Section 1R21.5.b(2))05000454/2015008-06; 05000455/2015008-06

NCV

Failure to Evaluate the Adverse Effects of Changing the

SXCT Tornado Analysis as Described in the UFSAR

(Section 1R21.5.b(3))05000454/2015008-07; 05000455/2015008-07

FIN

Failure to Evaluate the Adverse Effects of Changing the

SXCT Tornado Analysis as Described in the UFSAR

(Section 1R21.5.b(3))05000454/2015008-08; 05000455/2015008-08

NCV

Failure to Provide Proper Direction for Low Level

Isolation of the RWST in EOPs (Section 1R21.6.b(1))05000454/2015008-09; 05000455/2015008-09

VIO

Failure to Promptly Correct an NRC-Identified NCV

Associated with the Capability to Detect and Isolate

ECCS Leakage (Section 4OA2.1.b(1))

2 05000454/2015008-10; 05000455/2015008-10

NCV

Failure to Maintain the Instrument Loops Used to Verify

Compliance with the Containment Average Air

Temperature TS Limit (Section 4OA2.1.b(2))05000454/2015008-11; 05000455/2015008-11

NCV

Operability Evaluation Relied on Probabilities of

Occurrence of the Associated Event

(Section 4OA2.1.b(3))

Closed 05000454/2015008-03; 05000455/2015008-03

NCV

Failure to Evaluate the Adverse Effects of Sharing the

RWSTs of Both Reactor Units (Section 1R21.5.b(1))05000454/2015008-04; 05000455/2015008-04

FIN

Failure to Evaluate the Adverse Effects of Sharing the

RWSTs of Both Reactor Units (Section 1R21.5.b(1))05000454/2015008-05; 05000455/2015008-05

NCV

Failure to Adequately Implement a Design Change

Associated with the RWSTs (Section 1R21.5.b(2))05000454/2015008-06; 05000455/2015008-06

NCV

Failure to Evaluate the Adverse Effects of Changing the

SXCT Tornado Analysis as Described in the UFSAR

(Section 1R21.5.b(3))05000454/2015008-07; 05000455/2015008-07

FIN

Failure to Evaluate the Adverse Effects of Changing the

SXCT Tornado Analysis as Described in the UFSAR

(Section 1R21.5.b(3))05000454/2015008-08; 05000455/2015008-08

NCV

Failure to Provide Proper Direction for Low Level

Isolation of the RWST in EOPs (Section 1R21.6.b(1))05000454/2015008-10; 05000455/2015008-10

NCV

Failure to Maintain the Instrument Loops Used to Verify

Compliance with the Containment Average Air

Temperature TS Limit (Section 4OA2.1.b(2))05000454/2015008-11; 05000455/2015008-11

NCV

Operability Evaluation Relied on Probabilities of

Occurrence of the Associated Event

(Section 4OA2.1.b(3))

3

LIST OF DOCUMENTS REVIEWED

The following is a list of documents reviewed during the inspection. Inclusion on this list does

not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that

selected sections of portions of the documents were evaluated as part of the overall inspection

effort. Inclusion of a document on this list does not imply NRC acceptance of the document or

any part of it, unless this is stated in the body of the inspection report.

CALCULATIONS

Number

Description or Title

Revision

4391/19D-11

Sizing of Replacement Battery Charger for Diesel Driven

Pumps

0

BYR08-035

Essential Service Water Cooling Tower Basin Level

Indication Uncertainty Analysis

0

BYR12-070

Auxiliary Building Environment following a High Energy Line

Break in the Turbine Building

2

BYR12-072

Thermal Endurance Evaluation of the Safety Related

Electrical Equipment in the Essential Service Water (SX)

Cooling Tower Switchgear Rooms

0

BYR97-193

Battery Duty Cycle and Sizing for the Byron Diesel Driven

Auxiliary Feedwater Pumps and the Byron Diesel Driven

Essential Service Water Makeup Pumps

1-1E

BYR97-205

125VDC Battery Charger Sizing Calculation

2

BYR97-204

125 VDC Battery Sizing Calculation

3-3K

BYR97-224

125Vdc Voltage Drop Calculation

4-4A

BYR97-226

125 V DC System Short Circuit Calculation

4

BYR97-239

SX Cooling Tower Basin Level Auto Start Level Set Point

Analysis

1

BYR97-336

SX Cooling Tower Basin - Time to Reach the Low Level

Alarm Set Point

1

BYR2000-136

Voltage Drop Calculation for 4160V Switchgear Breaker

Control Circuits

1

BYR2000-191

Voltage Drop Calculation for 480V Switchgear Breaker

Control Circuits

0 -0C

4391/19-AN-3

Protective Relay Settings for 4.16 kV ESF Switchgear

16

19-AQ-24

Voltage Drop on 480-120V AC Control Transformer Circuits

8

19-AQ-63

Division Specific Degraded Voltage Analysis

7A

19-AQ-69

Evaluation of the Adequacy of the 120 Vac Distribution

Circuit at the Degraded Voltage Setpoint

16

19-AQ-75

Essential Service Water Cooling Tower 480V Buses

Maximum Voltage

1

19-AU-4

480 V Unit Substation Breaker and Relay Settings

19

19-G-1

Cable Ampacity

2

19-T-5

Diesel Generator Loading During LOOP/LOCA

7

BYR01-068

Environmental Parameters of EQ Zones

2

BYR01-084

Generic Thermal Overload Heater Sizing Calculation for

Motor Operated Valves

000

4

CALCULATIONS

Number

Description or Title

Revision

BYR01-095

Motor Operated Valves (MOV) Actuator Motor Terminal

Voltage and Thermal Overload Sizing Calculation - Essential

Service Water (SX) System

1

BYR06-111

Model APT-30K-11 SXCT Fan Blade Pitch Setting

1

BYR12-042

Essential Service Water Discharge Header Temperature

Indication Uncertainty

0

BYR95-005

120 VAC Instrument Bus/SSPS Cabinet Fuse Sizing and

Coordination

0

BYR96-128

Refueling Water Storage Tank (RWST) Level Alarm

Bistables and Level Indication Accuracy

2

DIT BB-EPED-

0189

Design Information Transmittal: Minimum Starting/Running

Voltages for Essential Motors

5/14/93

DIT BB-EXT-

0406

Design Information Transmittal: Essential Service Water

Cooling Tower Fan Motors [starting duty]

12/9/92

DIT-BRW-2002-

033

Design Information Transmittal: Basis for EDG loading

10/15/02

SI-90-01

Minimum Containment Flood Level

11

BYR04-016

RHR, SI, CV, and CS Pump NPSH During ECCS Injection

Mode

2

BYR14-053

Pressurizer PORV Air Accumulator Tank Requirements

0

BYR06-029

Byron/Braidwood SI/RHR/CS/CV system hydraulic analysis

in support of GSI-191

5

BYR06-058

NPSHA for RHR & CS Pumps During Post-LOCA

Recirculation

0

BYR07-055

Determination of the Correlation for the Critical Submergence

Height (Vortexing) for the RWST

0

SM-SI0930

RWST Level

D

SITH-1

Refueling Water Storage Tank (RWST) Level Set points

8

CN-RRA-00-47

Byron/Braidwood Natural Circulation Cooldown TREAT

Analysis for RSG and Uprating Programm

3

CN-RRA-00-47

Byron/Braidwood Natural Circulation Cooldown TREAT

Analysis for RSG and Uprating Program

4

CQD-200074

PORV Accumulator Tank

Z2

8.1.16

Refueling Water Storage Tanks Analysis and Design

5

BYR97-287

Determination of RWST Free Air Volume above Maximum

RWST Water Level

2

SM-SI0930

RWST Level

D

SM-SI0931

RWST Level

D

SM-SI0932

RWST Level

D

SM-SI0933

RWST Level

D

ATD-0062

Heat Load to the Ultimate Heat Sink During a Loss of

Coolant Accident

5

BYR03-131

Evaluation of UHS Make Up for CST-based Cooldown Profile 1

BYR05-018

Tornado Missile Risk Assessment of Vulnerable Targets of

Essential Service Water Cooling Towers

0

BYR06-111

Model APT-30K-11 SXCT Fan Blade Pitch Setting

1

5

CALCULATIONS

Number

Description or Title

Revision

BYR09-002

UHS Capability with Loss of SX Fans due to a Tornado

Event

1

BYR09-002

UHS Capability with Loss of SX Fans due to a Tornado

Event

1

BYR97*239

SX Cooling Tower Basin Level Auto Start Setpoint Error

Analysis

1

BYR97-034

Essential Service Water Cooling Tower Basin Minimum

Volume Versus Level and Minimum

Usable Volume Calculation

0a

BYR97-034

Essential Service Water Cooling Tower Basin Minimum

Volume Versus Level and Minimum

Usable Volume Calculation

0A

BYR97-127

Byron Ultimate Heat Sink Cooling Tower Performance

Calculations

1

BYR97-134

Heat Load on the UHS - 2 Unit Shutdown

3

BYR97-366

SX Cooling Tower Basin - Time to Reach the Low Level

Alarm Set Point

1

BYRO8-035

Essential Service Water Cooling Tower Basin Level

Indication Uncertainty Analysis

0

NED-M-MSD-

009

Byron Ultimate Heat Sink Cooling Tower Basin Temperature

Calculation: Part IV

8B

NED-M-MSD-

014

Byron Ultimate Heat Sink Cooling Tower Basin Makeup

Calculation

9

UHS-01

Ultimate Heat Sink Design Basis LOCA Single Failure

Scenarios

4

SL-101

ELMS-AC Report: Running Voltage Summary, Division 12

1/21/15

SL-102

ELMS-AC Report: Short Circuit Summary for High Voltage

Buses

1/21/15

SL-109

ELMS-AC Report: Connection Loading, Division 12

1/21/15

SL-112

ELMS-AC Report: Single Bus Summary, Bus 142

4/20/15

CORRECTIVE ACTION DOCUMENTS Generated Due to the Inspection

Number

Description or Title

Date

AR02488878

2015 CDBI - Design Analysis Inconsistency Identified

4/21/15

AR02489108

NRC CDBI: Loose Parts Found During Walkdown of RWST

4/22/15

AR02489149

CDBI - Bucket Collecting Diesel Fuel Drips from 0DO088A

4/22/15

AR02489198

CDBI - SX Make-Up Pump Temperature Recorder Panel

Memory Full

4/22/15

AR02489297

CDBI - Outdated Information in SystemIQ

4/22/15

AR02489456

NRC ID: Jumpers Not Readily Available for 1/2BOA PRI-5

4/22/15

AR02489360

Negative Vibration Reading on Idle 0E SXCT Fan

4/22/15

AR02490324

CDBI - ID 1RY456 WO As-Found Not as Expected, No IR

Written

4/24/15

AR02493191

CDBI - Issues Identified in Calculation BYR 97-224

4/30/15

AR02493990

CDBI - Issue Identified in Calculation 19-AQ-69

5/1/15

AR02495580

CDBI Question Related to BEP ES-1.3 Cold Leg

Recirculation

5/4/15

6

CORRECTIVE ACTION DOCUMENTS Generated Due to the Inspection

Number

Description or Title

Date

AR02495584

CDBI - FC Purification Flow Not Considered in RWST NPSH

Calc

5/4/15

AR02495866

CDBI - NRC Identified Issues in BYR97-193

5/5/15

AR02496142

CDBI - 50.59 and DRP did not explicitly evaluate GDC 5

5/5/15

AR02495973

NRC CDBI - Error Discovered in EACE Investigation

5/6/15

AR02496766

CDBI - RWST Calc May Lead to Inconsistent Application of

TS

5/6/15

AR02497347

NRC CDBI: Procedure Enhancement for ECCS Flow

Balancing

5/6/15

AR02497940

CDBI Deficiency Identified - THD Testing for Instrument

Inverter

5/8/15

AR02497925

Lightning Rod on SX Cooling Tower Bent; Clarify Inspection

WO Instructions

5/8/15

AR02501392

CDBI 2015 - VTIP for Containment DP Has Limited Lead

Length

5/15/15

AR02501454

CDBI - CA Created for NCV Does Not Resolve Issue

5/15/15

AR02502846

No Routine PM on Containment Temperature Loops

5/19/15

AR02504624

CDBI Concern Regarding Op Eval 13-007

5/22/15

AR02504475

CDBI - TS Clarification Needed for Transition to LTOPs

5/22/15

AR02506214

2012 50.59 for SXCT Tornado Analysis

5/19/15

CORRECTIVE ACTION DOCUMENTS Reviewed During the Inspection

Number

Description or Title

Date

AR00301744

Design of RWST Vacuum Relief System

2/15/05

AR00239280

RWST Vent / Vacuum Breaker Design Basis Issues

7/27/04

AR00880223

0A SX M/U PP Failures

2/13/09

AR00881611

0A SX MU Pump Did Not Stop When Local CS Taken to Off

2/17/09

AR01053940

1DC08E Battery, 1DC08E 123 Bus and DC 123 Batt Low

4/8/10

AR01115570

DC Bus 123 Low Voltage

9/21/10

AR01204963

Megger Test of Submerged Cable (1SX172)

4/20/11

AR01217212

Check/Adjust Charger 123 Float Voltage

5/17/11

AR01263407

0A SX MU PP Failed to Start at the Desired Setpoint SPC

9/15/11

AR01318043

0A SX M/U PP Battery Bank Test

1/25/12

AR01362643

Replace Breaker for MCC 035-2-C5 (0CW03PC-C)

5/4/12

AR01368220

CDBI ESF MCC Contactors not Tested at Assumed Pickup

Volt

5/18/12

AR01376793

CDBI Follow-up on MCC Contactors (IR 1368220)

6/11/12

AR01377764

NRC CDBI - Protective Relay Setting Tolerances

6/12/12

AR01378259

Need Engineering to Evaluate Test Frequency

6/15/12

AR01380744

Action Tracking Needed for Size 3 and 4 Contactors

6/22/12

AR01387518

The Station 111 ESF Battery Needs to Be Replaced in

B1R19

7/11/12

AR01387520

The Station 112 ESF Battery Needs to be Replaced in

B1R19

7/11/12

AR01390648

Protective Relay Tolerances Require Fleet Review

7/19/12

7

CORRECTIVE ACTION DOCUMENTS Reviewed During the Inspection

Number

Description or Title

Date

AR01398419

NRC IDD CDBI Green NCV Non-Conforming 480/120 VAC

Motor Contactors

6/15/12

AR01398426

NRC CDBI Green NCV Non-Conservative Cal Tolerance for

Elec Relays

6/15/12

AR01413695

Engineering Evaluate Frequency of Battery Capacity Test

9/16/12

AR01502583

0A SX Makeup Pump Failed to Auto Start per 0BOSR 7.9.6-

1

4/16/13

AR01518720

Breaker Will Not Reset During Oden Testing

5/29/13

AR01570572

0A SX M/U PP Had To Be Tripped During Monthly Run

10/10/13

AR01588590

Loss of Instrument Bus 111

11/21/13

AR01589264

Need New Contingency Work Order ofr Instrument Inverter

111

11/23/13

AR01590368

NRC ID - PCM Template/Vendor Manual Recommendation

11/26/13

AR01611287

0A SX Makeup Pump Auto Start Level Setpoint

1/23/14

AR01654589

Erratic Reading on Ammeter (111-IP001) for Inverter 111

4/30/14

AR01658463

Specific Gravity of Battery Cell Still Low After Equalize

5/10/14

AR01680303

0A SX MU PP Trouble Alarm Continues to Alarm

7/10/14

AR01693147

Gradual Float Current Trend on 111 Battery Charger

4/15/14

AR02407275

0SX02PA Kept Running

11/5/14

AR02417160

Pump As Found Condition/Dry Start Improvement

Opportunity

11/25/14

AR02440865

Thermography Needed on FRT for Instrument Inverter 111

11/29/14

AR02448283

0A SX MU Failed Surveillance

2/5/15

AR01299897

Replace Breaker for MCC 132Z1-A4 (0SX157A)

12/8/11

AR01056715

NER-NC-10-008-Y - Buried Cable

4/14/10

AR01322720

B2F26 Bus 142 Undervoltage Relay

2/3/12

AR01409309

Safety-Related Cable Vault 1M1G(1G1) Inspection - Repairs

9/5/12

AR01417720

MCC 132Z1-A5 Tripped Out of Tolerance

9/24/12

AR01425642

Safety-Related Cable Vault 1J2 Inspection - Repairs

10/12/12

AR01592242

Operating Experience Applicable to Byron (SXCT Fan

Reverse Rotation)

12/2/13

AR01625774

Degraded Voltage Relay Target did not Change State

2/25/14

AR01648079

Step Change Identified in Unit 1 Containment Air

Temperature in PI

4/16/14

AR01687277

Safety Related Cable Vault PM and Engineering Inspections

7/30/14

AR02437410

Cable Vault PM and Engineering Inspections

1/14/15

AR02437973

CDBI FASA - Review of Robinson and Wolf Creek Findings

1/15/15

AR00239280

RWST Vent/Vacuum Breaker Design Basis Issue

7/27/04

AR01360789

U-1 RWST level

4/30/12

AR01361308

U-1 RWST on FC Purification

5/2/12

AR01361838

U-1 RWST level loss During Purification

5/3/12

AR0128230

NRC Information Notice 2012-01: Seismic Considerations -

Principally Issues Involving Tanks

5/9/12

AR01398434

NRC CDBI Green NCV-Leak Detection for ECCS Flowpath

Lacking

6/15/12

AR01378257

CDBI, Question about ECCS leakage

6/15/12

8

CORRECTIVE ACTION DOCUMENTS Reviewed During the Inspection

Number

Description or Title

Date

AR01465872

Review of Braidwood IR 1459353 Pzr PORV Accumlator

Press

1/23/13

AR01635829

1B PZR PORV Accum Failed Decay Test

3/19/14

AR02454767

NOS ID: No CA to Correct an NRC NCV

2/18/15

IR298958

SSD&PC: Inaccurate Setpoints Referenced in BYR97-034

6/30/05

AR 01546621

Inadequate 50.59 for EC 385829 (SXCT Tornado Missiles)

8/14/13

AR295141295141

Ssd&pc Question on Tornado Anaylsis Supporting UFSAR

Stmnt

1/28/05

AR1677584

Clarification Needed on UHS Passive Failure Design

7/1/14

AR1567903

NRC Question and Feedback on UHS Temperature Analysis 10/3/13

AR1677513

UFSAR Section 2.4.11.6 Needs Revision

7/1/14

AR1677646

Recommendation from UHS Assessment

7/1/14

AR1546621

Inadequate 50.59 for EC 385829

2/9/12

AR2406579

Failed Spider Bearing on 0A SX Makeup Pump

11/4/14

AR1269014

Obsolete SX Makeup Pump D/O Storage Tank Level

Indicator

9/28/11

AR2437508

Review of Flow Anomaly On 0B SX Makeup

1/14/15

AR2448283

0A SX MU Failed Surveillance

2/5/15

DRAWINGS

Number

Description or Title

Revision

S-529

Essential Service Cooling Tower Drainage Duct Plan,

Section Details

H

6E-0-4030SX09

Schematic Diagram - Essential Service Water Make-up

Pump 0A 0SX02PA

P

6E-0-4030SX23

Schematic Diagram - Essential Service Water Make-up

Pump 0A Control Cabinet (Diesel Driven) 0SX02JA

S

6E-0-4030SX24

Schematic Diagram - Essential Service Water Make-up

Pump 0A Control Cabinet (Diesel Driven) 0SX02JA

Annunciator

F

6E-0-

4030CW11

Schematic Diagram - Essential service Water Cooling Tower

0A & 0B Well Water Make-up Valves 0CW100A & B

D

6E-0-

4030WW01

Schematic Diagram - Deep Well Pump 0A - 0WW01PA

M

6E-0-

4030WW02

Schematic Diagram - Deep Well Pump 0B - 0WW01PB

H

6E-0-

4030WW05

Schematic Diagram - Essential service Water Cooling Tower

0A & 0B Circulating Water Make-up Valves 0WW019A & B

E

6E-1-4001A

Station One Line Diagram

P

6E-1-4001E

Station Key Diagram

O

6E-1-4002E

Single Line Diagram - 120V AC ESF Instrument Inverter Bus

111 and 113, 125V DC ESF Distribution Center 111

K

6E-1-4007A

Byron - Unit 1 - Key Diagram 480V ESF Substation Bus

131X (1AP10E)

M

6E-1-4010A

Key Diagram - 125V DC ESF Distribution Center Bus 111

(1DC05E) Part 1

M

9

DRAWINGS

Number

Description or Title

Revision

6E-1-4010B

Key Diagram - 125V DC ESF Distribution Center Bus 111

(1DC05E) Part 2

G

6E-1-4010C

Key Diagram - 125V DC Non Safety Related Distribution

Panel 113 (1DC05EB)

K

6E-1-4030DC05

Schematic Diagram - 125 VDC ESF Distribution Center, Bus

111, Part 1, 1DC05E

U

6E-1-4030IP01

Schematic Diagram 7.5KVA Fixed Frequency Inverter for

Instrument Bus 111 (1IP05E)

0

6E-1-4030RC31

Schematic Diagram - Reactor Coolant System High Pressure

& Low Temperature Control & Alarms

G

6E-1-4030RH02

Schematic Diagram - Residual Heat Removal Pump 1B -

1RH01PB

N

6E-1-4030RY14

Schematic Diagram - Pressurizer Pressure & Level Control

Safety Related & Non-Safety Related (Div 12)

F

6E-1-4030RY17

Schematic Diagram - Pressurizer Power Relief Valves -

1RY455A & 1RY456; Pressurizer Relief Tank Primary Water

Supply Isolation Valve - 1RY8030; Pressurizer Relief Tank

Drain Isolation Valve 1RY8031

V

6E-1-4031RC26

Loop Schematic Diagram - Reactor Coolant System Cold

Overpressurization System Control 1A & 1D Control Cabinet

5 & 6

S

6E-1-4031RY15

Loop Schematic Diagram - Pressurizer Pressure & Level

Control Cabinet 6 (1PA06J) Part 1

O

6E-1-4031RY19

Loop Schematic Diagram - Pressurizer Pressure Safety

Valve Discharge Temp & Pressure Control (ITE-0464)

Control Cabinet 7 (1PA07J)

F

M-42 Sh. 6

Diagram of Essential Service Water

BC

M-60 Sh. 5

Diagram of Reactor Coolant

AO

M-2042 Sh. 5

P&ID/C&I Diagram ESS Service Water System - SX

F

6E-0-1003

Duct Runs, Outdoor Plan, Southeast Area

AC

6E-0-1004

Duct Runs, Outdoor Plan, Southwest Area

Y

6E-0-1009

Duct Runs, Sections

F

6E-0-3502

Electrical Installation, ESW Cooling Tower 0A Plan -

Switchgear Room, Elev. 874-6

AZ

6E-0-3502CT1

Conduit Tabulation, ESW Cooling Tower 0A Plan -

Switchgear Room, Elev. 874-6

T

6E-0-3502D01

Electrical Installation, ESW Cooling Tower 0A Switchgear

Room Partial Plans and Sections

N

6E-0-3507

Electrical Installation, ESW Cooling Tower 0B Plan -

Switchgear Room, Elev. 874-6

BN

6E-0-3507CT1

Conduit Tabulation, ESW Cooling Tower 0B Plan -

Switchgear Room, Elev. 874-6

Y

6E-0-3507D01

Electrical Installation, ESW Cooling Tower 0B Switchgear

Room Partial Plans and Sections

W

6E-0-4030SX01

Schematic Diagram, Essential Service Water Cooling Tower

0A, Fan 0A

V

6E-0-3680

Duct Run Routing Outdoor - West of Station

AC

10

DRAWINGS

Number

Description or Title

Revision

6E-0-4030SX02

Schematic Diagram, Essential Service Water Cooling Tower

0A, Fan 0B

U

6E-0-4030SX03

Schematic Diagram, Essential Service Water Cooling Tower

0A, Fan 0C

U

6E-0-4030SX04

Schematic Diagram, Essential Service Water Cooling Tower

0A, Fan 0D

W

6E-0-4030SX05

Schematic Diagram, Essential Service Water Cooling Tower

0B, Fan 0E

V

6E-0-4030SX06

Schematic Diagram, Essential Service Water Cooling Tower

0B, Fan 0F

W

6E-0-4030SX07

Schematic Diagram, Essential Service Water Cooling Tower

0B, Fan 0G

W

6E-0-4030SX08

Schematic Diagram, Essential Service Water Cooling Tower

0B, Fan 0H

W

6E-1-4001A

Station One Line Diagram

P

6E-1-4006B

Key Diagram, 4160V ESF Switchgear Bus 142

J

6E-1-4008AN

Key Diagram, 480V ESW Cooling Tower ESF MCC 132Z1

R

6E-1-4012A

Key Diagram, 120 Vac Instrument Bus 111

W

6E-1-4018B

Relaying & Metering Diagram, 4160 ESF Switchgear Bus

142

U

6E-1-

4030AP115

Schematic Diagram, Tripping Circuit, 480V ESW Cooling

Tower MCC 131Z1A, 132Z1A

A

6E-1-4030RY17

Schematic Diagram, Pressurizer Power Relief Valve 1RV456

V

6E-1-4030SI02

Schematic Diagram, Safety Injection Pump 1B

N

6E-1-4030SI14

Schematic Diagram, Containment Sumps 1A and 1B

Isolation Valves SI8811A & B

Q

6E-1-4031VP11

Loop Schematic Diagram [containment inside/outside

differential pressure]

K

M-61, Sh. 1B

Diagram of Safety Injection

AX

M-136, Sh. 1

Diagram of Safety Injection

BB

M-63, Sh. 1A

Diagram of Fuel Pool Cooling and Clean up

BI

S-1404

Refueling Water Storage Tank Sections & Details

I

M-60, Sh. 8

Diagram of Reactor Coolant

AA

98Z512-001-2,

Sh. 1

Pressurizer PORV Air Relief Valve

0

M-60, Sh.5

Diagram of Reactor Coolant

AO

10 CFR 50.59 DOCUMENTS (Screenings/Safety Evaluations)

Number

Description or Title

Date

6G-97-0110

DCP 9600355 ESW Cooling Tower Basin Level Switch

7/3/97

EC385829

Tornado Missile Design Basis for the Essential Service

Water Cooling Tower

0

6G-11-004

Tornado Missile Design Basis for the Essential Service

Water Cooling Towers

2/9/12

EC385951

Multiple Spurious Operation - Scenario 14, 1SI8811A/B

12/9/11

6E-05-0172

UFSAR Change Package (DRP)11-052

11/16/05

11

10 CFR 50.59 DOCUMENTS (Screenings/Safety Evaluations)

Number

Description or Title

Date

6E-15-035

Increase Pressurizer PORV tank Operating Pressure to

Increase Margin for PORV Operation (Unit 1)

0

6H-00-0155

Technical Requirements Manual (TRM) Revision to Delete

TLCO 3.4.a, Pressurizer Safety Valves-Shutdown

9/19/00

MISCELLANEOUS

Number

Description or Title

Date or

Revision

IST Program Plan - Service Water System

8/26/14

Standing Order 15-020

Emergency Operating Procedure Cold Leg Recirc.

5/15/15

DW-09-004

ERG Feedback

2/27/09

Stewart & Stevenson Certificate of Conformance for Battery

Chargers Serial No. 2165, 2167, 2170, 2174, 4 Batterrie 20

Cells/Set and 8 Battery Racks, Purchase Order No. 203731

11/4/81

06EN003246

FLT Series Flex Switch - Flow, Level, Temperature Switch

Monitor

2

01492090-03

Level 3 OPEX Evaluation - NRC IN 2013-05: Battery

Expected Life and Its Potential Impact on Surveillance

Requirements

5/16/13

CQD-009436

Seismic Qualification Test Report for Nife Ni-Cad Batteries

H-410 (1,2 AF01EA-A, EA-B, EB-A, EB-B/0SX02EA, EB-A,

EC-A, ED-A

8/17/83

CQD-012527

Review of Seismic Qualification Test Report for Battery

Chargers (1&2 DC03E, 04E)

10/2/13

CQD-049161

Justification for the Application of Permatex Form A Gasket

with EPT Diaphragms

1

CQD-200164

Dynamic Qualification of Battery Chargers 0SX02EA-1

through 0SX02ED-1; 1,2AF01EA-1 and 1,2AF01EB-1

5/29/86

NEC-06-6066

Procurement of Safety Related 125 Volt Batteries

B

604990-70-F1

Reliance Electric Dimension Sheet [SX Cooling tower fan

motor data sheet]

4/4/78

EQ-GEN023

EQ Binder for NAMCO EA180 limit switches

13

EC-397415

EQ Evaluation - Pressurizer PORV Diaphragm Design

Pressure

0

EQER-06-98-

002

EQ Evaluation for PORVs 1(2) FSV-RY-455A & 1(2)FSV-RY-

456

2/29/99

Low Temperature Protection (LTOP) System Evaluation for

Byron and Braidwood Units 1 and 2 Measurement

Uncertainty Recapture (MUR) Power Uprate Program

9/7/10

Simulator Work

Request 13961

PZR PORV Testing reveals lower than design flow

4/25/12

Byron Unit 1 Pressure and Temperature Limits Report

3/14

EC 381986

Summary of the Design and Licensing Basis for Inadvertent

ECCS Actuation at Power

0

12

MODIFICATIONS

Number

Description or Title

Date or

Revision

EC394865

Ultimate Heat Sink Capability with Loss of Essential Service

Water Cooling Tower Fans

2

EC385829

UHS Capability with Loss of SX Fans Due to Tornado

Missiles

2/14/12

M6-1(2)-87-142

Install Fan Cooling to Instrument Power Inverter Cubicles

10/17/90

EC385951

Multiple Spurious Operation - Scenario 14, 1SI8811A/B

12/9/11

EC388735

Detailed Review of FC Purification System for Use of Non

Safety Related Portion Connected to Safety Related Piping

0

EC396016

Increase U1 Pressurizer PORV Accumulator Tank Operating

Pressure to Increase number of PORV Open/Close Cycles

from Accumulator

0

OPERABILITY EVALUATIONS

Number

Description or Title

Date 13-001

Capacity of the Pressurizer PORV Air Accumulator During

Natural Circulation Cooldown

5 13-007

Ultimate Heat Sink Capability with Loss of Essential Service

Water Cooling Tower Fans

1

PROCEDURES

Number

Description or Title

Revision

1BOA PRI-5

Control Room Inaccessibility

108

1BOA ELEC-5

Local Emergency Control of Safe Shutdown Equipment

106

0BOA PRI-7

Loss of Ultimate Heat Sink Unit 0

1

1BOA PRI-7

Essential Service Water Malfunction Unit 1

106

1BEP ES-1.3

Transfer to Cold Leg Recirculation Unit 1

204

1BCA-1.2

LOCA Outside Containment Unit 1

200

OP-AA-102-106

Operator Response Time Program

3

OP-BY-102-106

Operator Response Time Program at Byron Station

7

1BOA S/D-2

Shutdown LOCA Unit 1

105

1BOSR XRS-Q1 Unit One Remote Shutdown Panel Quarterly Surveillance

13

1BFR-H1

Response to Loss of Secondary Heat Sink Unit1

203

0BHSR 8.4.2-1

Unit Zero Comprehensive Inservice Testing (IST)

Requirements for Essential Service Water Makeup Pump 0A

8

0BHSR SX-1

Unit 0 0A SX Makeup Pump Battery Bank A Capacity Test

0

0BHSR SX-5

0A SX Makeup Pump Battery Bank D Capacity Test

0

0BISR 7.a.4-200 Calibration of Essential Service Water Cooling Tower Basin

0A Level Switch (SX)

7

0BOSR Z.7.a.2-

1

Unit Common Deepwell Pump Operability Monthly

Surveillance

1

0BOSR 7.9.6-1

Essential Service Water Makeup Pump 0A Monthly

Operability Surveillance

32

0BVSR SX-1

Unit 0 0A SX Makeup Pump Battery Bank A Capacity Test

3

0BVSR SX-4

Unit 0 0A SX Makeup Pump Battery Bank D Capacity Test

3

0BVSR WW-1

Biennial Deep Well Pump Structure Inspection

2

1BHSR 8.4.2-1

Unit 1 Bus 111 125V Battery Charger Operability

1

13

PROCEDURES

Number

Description or Title

Revision

1BHSR 8.4.3-1

Unit 1 125 Volt Battery Bank 111 Service Test

3

1BHSR 8.6.6-1

Unit 1 Battery 111 125 Volt Battery Bank 5 Year Modified

Performance Test

0 & 2

1BHSR AF-1AA

Unit 1 1B Diesel Aux Feed Pump Battery Bank A Battery A

(1AF01EA-A) Capacity Test

1

1BOA ELEC-1

Loss of DC Bus Unit 1

103

1BOSR 8.4-1

125V DC Bus 111 Load Shed When Cross-Tied to DC Bus

211

12

2BHSR 8.4.2-1

Unit 2 Bus 211 125V Battery Charger Operability

1

BISR 3.1.10-206 Pressurizer Pressure Protection Channel II (RY) Test Report

Package)

8

BISR 3.1.10-207 Pressurizer Pressure Protection Channel III (RY) Test Report

Package)

8

BISR 4.12.8-200 Wide Range Reactor Coolant Pressure Loop 1A Hot Leg

(RC)

7

BOP-AP-93

MCC 035-2 Outage

1

BOP SX-3

Essential Service Water Make-up Pump Startup

30

BOP SX17

Shutdown of SX Makeup Pump Battery Chargers

3

BOP SX18

Placing the SX Makeup Pump Battery Chargers in

Operation/Equalize

8

CC-AA-308

Control and Tracking of Electrical Load Changes

4

ER-AA-310-

1004

Maintenance Rule - Performance Monitoring

13

MA-BY-026-

1001

Seismic Housekeeping

2

MA-BY-721-060

125 Volt Battery Bank 18 Month Surveillance

11

MA-BY-721-061

125 Volt Battery Bank Quarterly Surveillance

12 & 15

MA-BY-723-053

Station Battery Charger 18 Month Surveillance

18

MA-BY-723-

053-001

0B SX Makeup Pump A Battery Charger 0SX02EA Battery

Charger Test

0

MA-BY-723-

053-002

0B SX Makeup Pump D Battery Charger 0SX02ED Battery

Charger Test

1

MA-BY-723-

053-003

0B SX Makeup Pump B Battery Charger 0SX02EB Battery

Charger Test

0

MA-BY-723-

053-004

0B SX Makeup Pump C Battery Charger 0SX02EC Battery

Charger Test

1

MA-BY-723-054

Nickel Cadmium Battery Bank Surveillance

14

0BHSR SX-3

Annual Surveillance for Essential Service Water Cooling

Tower Fan Motors

2

0BOSR 7.9.4-1

ESW Cooling Tower Fan Monthly Surveillance

6

1BOSR IP-R1

Instructions to Cycle Instrument Bus 111 Distribution Panel

Molded Case Circuit Breakers

0

1BOSR 3.2.9-1

Train A Manual Safety Injection Initiation and Manual Phase

A Initiation Surveillance

22

1BOSR 8.9.1-2

Unit 1 ESF Onsite Power Distribution Weekly Surveillance

Division 12

10

BOP MP-19

Adjusting Reactive Load

12

14

PROCEDURES

Number

Description or Title

Revision

ER-AA-300-150

Cable Condition Monitoring Program

1

MA-AA-723-330

Electrical Testing of AC Motors Using Baker Instrument

Advanced Winding Analyzer

3

MA-AA-725-102

Preventative Maintenance on Westinghouse Type DHP 4kv,

6.9kv, and 13.8kv Circuit Breakers

8

1BGP-100-5

Plant Shutdown and Cooldown

68

BOP FC-7

Startup of the Purification System to Purify or Recirculate the

Refueling Water Storage Tank

13

1BEP ES-0.2

Natural Circulation Cooldown Unit 1

202

BAR 1-12-C4

RCS Press High at Low Temp

2

1BOSR 5.C.3.1

Safety Injection System Cold Leg Flow Balance

3

2BOSR 0.1-4

Unit 2 Mode 4 Shiftly and Daily Operating Surveillance

25

1BOSR 0.1-

1,2,3

Unit 1 Mode 1,2,3 TRM and Tech Spec and Non Tech Spec

Data Sheet D5

56

BIP 2500-088

Calibration of Refueling Water Storage Tank Outlet

Temperature Loop (SI)

5

1BOSR

5.5.8.SI.5-2C

Unit 1 Comprehensive Inservice Testing (IST) Requirements

for Safety Injection Pump 1SI01PB

5

1BOSR

5.5.8.SI.5-2a

Unit 1 Group A Inservice Testing (IST) Requirements for

Safty Injection Pumps 1SI01PB

1

0BOSR NLO-

TRM

Non-Licensed Operator TRM, ISFSI, and NPDES Data Daily

Logs

18

1BGP 100-5

Plant Shutdown and Cooldown

68

BOP SX-T2

SX Basin Level Tree

5

BOP SX-11

SXCT Fan Startup

9

BOP SX-12

Makeup to an Essential Service Water Mechanical Draft

Cooling Tower

10

0BOA ENV-1

Adverse Weather Conditions

114

1BOA PRI-5

Control Room Inaccessibility

108

1BOA ELEC-5

Local Emergency Control of Safe Shutdown Equipment Unit

1

106

1BEP-1

Reactor Trip or Safety Injection

207

1BEP ES-0.1

Reactor Trip Response

203

1BEP ES-0.2

Natural Circulation Cooldown

202

BOP RH-6

Operation of the RH System In Shutdown Cooling

46

OP-AA-108

Oversight and and Control of Operator Burdens

2

BOP CC-1

Component Cooling Water System Startup

12

SURVEILLANCES (Completed)

Number

Description or Title

Date or

Revision

0BHSR SX-1

0A SX Makeup Pump Battery Bank A Capacity Test

6/14/12

0BHSR SX-5

0A SX Makeup Pump Battery Bank D Capacity Test

9/14/12

0BISR 7.a.4-200 Calibration of Essential Service Water Cooling Tower Basin

0A Level Switch (SX)

8/7/14

0BOSR

5.5.8.SX.5-1c

0SX02PA Comprehensive IST Req for SX Makeup Pump

2/5/15

15

SURVEILLANCES (Completed)

Number

Description or Title

Date or

Revision

0BOSR 7.9.6-1

0A SX Makeup Pump Operability Surveillance

3/12/13

0BOSR 7.9.6-1

0A SX Makeup Pump Operability Surveillance

2/4/15

0BOSR 7.9.6-1

0A SX Makeup Pump Battery Bank A Capacity Test

3/11/15

0BVSR SX-1

0A SX Makeup Pump Battery Bank A Capacity Test

10/17/06

0BVSR SX-4

0A SX Makeup Pump Battery Bank D Capacity Test

6/19/06

1BHSR 8.4.2-1

Unit 1 Bus 111 125V Battery Charger Operability Test

11/8/11

1BHSR 8.4.2-1

Unit 1 Bus 111 125V Battery Charger Operability Test

9/17/13

1BHSR 8.4.3-1

111 A Train 125V Battery Bank Service Test

3/20/14

1BHSR 8.6.6-1

111 A Train 125V Battery Bank 5Yr Capacity Test

4/1/08

1BHSR 8.6.6-1

111 A Train 125V Battery Bank 5Yr Capacity Test

9/11/12

BISR 3.1.10-206 Pressurize Pressure Protection Channel 2 Loop 1RY-0456

4/6/15

BISR 3.1.10-207 Pressurizer Pressure Protection Channel 3 Loop 1RY-0457

4/13/15

BISR 4.12.8-200 Cal of Wide Range RC Pressure Loop 1A Hot Leg 1P-406

4/28/14

M A-BY-721-

060

125 Volt Battery Bank Quarterly Surveillance

9/11/12

M A-BY-721-

060

125 Volt Battery Bank Quarterly Surveillance

3/20/14

M A-BY-721-

061

125 Volt Battery Bank 18 Months Surveillance

9/16/12

M A-BY-721-

061

125 Volt Battery Bank 18 Months Surveillance

3/22/14

M A-BY-721-

061

125 Volt Battery Bank 18 Months Surveillance

9/15/14

M A-BY-721-

061

125 Volt Battery Bank 18 Months Surveillance

12/16/14

MA-BY-723-053

EM 18 Month Battery Charger Surveillance - 0B SX M/U

Pump 0B Batt Chgr # 0SX02EB-1

1/15/13

MA-BY-723-053

EM 18 Month Battery Charger Surveillance - 0A SX M/U

Pump 0A Batt Chgr # 0SX02EA-1

2/6/14

MA-BY-723-053

EM 18 Month Battery Charger Surveillance - 0A SX M/U

Pump 0D Batt Chgr # 0SX02ED-1

8/5/14

MA-BY-723-053

EM 18 Month Battery Charger Surveillance - 0B SX M/U

Pump 0C Batt Chgr # 0SX02EC-1

3/27/15

MA-BY-723-

053-001

0B SX Makeup Pump A Battery Charger 0SX02EA Battery

Charger Test

2/4/14

MA-BY-723-

053-002

0B SX Makeup Pump D Battery Charger 0SX02ED Battery

Charger Test

8/6/14

MA-BY-723-

053-003

0B SX Makeup Pump B Battery Charger 0SX02EB Battery

Charger Test

1/15/13

MA-BY-723-

053-004

0B SX Makeup Pump B Battery Charger 0SX02EC Battery

Charger Test

3/27/15

MA-BY-723-054

Quarterly 24 VDC NiCad Battery Surveillance M/U Diesel

SX- 0SX02ED-A

8/5/14

MA-BY-723-054

Quarterly 24 VDC NiCad Battery Surveillance M/U Diesel

SX- 0SX02EA-A

9/5/14

16

SURVEILLANCES (Completed)

Number

Description or Title

Date or

Revision

MA-BY-723-054

Quarterly 24 VDC NiCad Battery Surveillance M/U Diesel

SX- 0SX02EA-A

10/30/14

MA-BY-723-054

NiCad Battery Surveillance M/U Diesel SX- 0SX02E

11/6/14

WO01579586

Unit 1 Pressurizer PORV Accumulator Press Decay Test

3/19/14

WO01774289

SI pump ECCS Flow Balance Test (After System Alteration)

10/5/14

WO01243123

OP 2BOSR 5.C.3-2 Unit 2 SI to HL Flow Balance

4/2/10

WO01243120

Unit 1 Safety Injection System Hot Leg Flow Balance

9/4/09

WO01243119

SI pump ECCS Flow Balance Test (After System Alterations)

9/4/09

WO01582134

1SI01PB Comprehensive IST RQMTS For Safety Injection

Pump

1/28/14

WO01425077

1SI01PB Comprehensive IST RQMTS For Safety Injection

Pump

8/9/12

WO01451296

STT/PIT For 1RY455A and 1RY456

9/28/12

WO01585186

STT/PIT For 1RY455A and 1RY456

2/7/14

PMID 140860

0BOSR 7.9.6-1 0A SX Makeup Pump Operability Review

4/18/13

TRAINING DOCUMENTS

Number

Description or Title

Date or

Revision

BY 14-2-2

Requalification Simulator Scenario Guide

1

10-1-5

Requalification Simulator Scenario Guide

0

P1-SPBY-1401

BEP-1, BEP-2

2

OPBYLLORT5

BFR H, Heat Sink Series

8/28/13

WORK DOCUMENTS

Number

Description or Title

Date or

Revision

00961518

Replace Entire Solenoid to Meet EQ Requirements - EM

ASCO Solenoid Valve Replacement (EQ) - 1FSV-RY456-2

4/1/08

01057719

Test All MCC Breakers in This MCC in a Bus Outage -

Assembly 480V RSH MCC 035-2

5/2813

01094421

Replace Float and Equalize Voltage Adjustment

Potentiometer

11/29/11

01490541

111 A Train 125 V Battery Charger Operability Test

9/18/13

01536066

Essential Service Water Cooling Tower Level 0SX-064 IM

Calibration

3/3/14

01558514

B1R19 Replace 111 ESF Batteries

3/29/14

01578627

Test Replace Actuator Hose 1RY456

3/14/14

01599481

Calibration of Wide Range RC Pressure Loop 1A Hot Leg

Pressure Loop 1RC-0406

4/28/14

01600072

Clean/Inspect/Check Connections on DC Bus/Panel 111 and

Perform Therm. on Distr. Panel Breakers

3/30/14

01621944

Support Diver Insp./Cleaning RSH South 0B Intake/SED PM

ID 30

6/25/13

01652815

211 A Train 125 V Battery Charger Operability Test

5/14/14

17

WORK DOCUMENTS

Number

Description or Title

Date or

Revision

01017127

Perform Dynamic Baker Testing - 1SI01PB Motor

8/26/08

01085998

Perform Static Baker Test and MA-AA-723-310 Inspection of

SX Cooling Tower Fan Motor 0SX03CC

4/27/09

01117942

PM for 4kV Bus 142, breaker ACB 1425Z

9/21/09

01119375

Lightning Protection System 5 Year Inspection [Includes

Document 1 attachment to WO]

11/18/09

01120491

PM for 4kV Bus 142, breaker ACB 1424

9/29/09

01129028

Inspection of SX Cooling Tower Fan Motor 0SX03D

10/28/09

01136617

PM for 4kV Bus 142, breaker ACB 1422

3/15/09

01141049

Perform Static Baker Test and MA-AA-723-310 Inspection of

SX Cooling Tower Fan Motor 0SX03CB

3/19/10

01216011

Perform Dynamic Baker Testing - 1SI01PB Motor

8/26/10

01258194

Calibration of OLS-XS097

1/6/11

01265167

PM for 4kV Bus 142, breaker ACB 1421

10/26/11

01287321

Inspection of SX Cooling Tower Fan Motor 0SX03CE

9/1/11

01299949

Containment Inside/Outside DP Loop 1VP-231

6/30/11

01343409

Inspection of SX Cooling Tower Fan Motor 0SX03CH

11/21/11

01367641

PM for 4kV Bus 142, breaker ACB 1SI01PB

2/21/12

01372340

PM for 4kV Bus 142, breaker ACB 1422

11/11/12

01382271

Perform Static Baker Test and MA-AA-723-310 Inspection of

SX Cooling Tower Fan Motor 0SX03CC

6/12/12

01384474-01

Inspection of SX Cooling Tower Fan Motor 0SX03CF

11/26/12

01380551-01

Inspection of SX Cooling Tower Fan Motor 0SX03CA

6/8/12

01393782

Inspection of SX Cooling Tower Fan Motor 0SX03CG

10/30/11

01401180

Calibration of OLS-XS097

8/24/12

01419437

PM for 4kV Bus 142, breaker ACB 1425Z

9/23/12

01419758

Test All MCC 132Z1 Breakers - Oden Testing

9/23/12

01420365

PM for 4kV Bus 142, breaker ACB 1424

9/23/12

01421751

Unit 1 Train A Manual SI and Manual Phase A Initiation

Surveillance

9/11/12

01433378-01

Inspection of SX Cooling Tower Fan Motor 0SX03CD

3/12/13

01453350

Containment Inside/Outside DP Loop 1VP-231

3/19/15

01471461

Calibration of OLS-XS096

9/6/11

01473594-01

Perform Static Baker Test and MA-AA-723-310 Inspection of

SX Cooling Tower Fan Motor 0SX03CB

5/17/13

01480666-01

Testing of Power Cables 2AP178

4/20/13

01486337

Calibration of OLS-XS096

2/8/13

01538412

PM for 4kV Bus 142, breaker ACB 1423

11/30/13

01564018-01

Testing of Power Cables 1AP178 (North SX towers)

3/18/14

01569220

Calibration of OLS-XS097

6/2/14

01585654-02

Testing of Power Cables 2AP183 (Bus 242, Cubicle 20)

10/6/14

01615167

Calibration of OLS-XS096

8/8/14

01621573-01

Perform Surveillance of SX Cooling Tower Fan Motor

0SX03CE

9/16/14

01639602

PM for 4kV Bus 142, breaker ACB 1421

11/19/14

18

WORK DOCUMENTS

Number

Description or Title

Date or

Revision

01644724-01

Perform Surveillance of SX Cooling Tower Fan Motor

0SX03CH

11/20/14

01652671

PM for 4kV Bus 142, breaker ACB 1SI01PB

3/29/15

01667453

Calibration of 1SX-015 Loop

2/17/15

01680518

Calibration of 1SX-016 Loop

3/31/15

01543156

Calibration of 2SX-015 Loop

2/12/14

01716477

Calibration of 2SX-016 Loop

3/23/15

01734645-01

SX Cooling Tower Fan Motor Surveillance - 0SX03CG

11/4/14

01734645-02

SX Cooling Tower Fan Motor Surveillance & Triannual

Inspection - 0SX03CG

11/5/14

01760801

PM for 4kV Bus 142, breaker ACB 1423

1/30/15

01805922

ESW Cooling Tower Fan Monthly Surveillance

3/10/15

01419750

Replace Actuator Diaphragm

9/20/12

01515448

Refueling Water Storage Tank Outlet Temp LOOP 1SI-058

2/24/14

01186461

Refueling Water Storage Tank Outlet Temp LOOP 1SI-058

4/21/10

01544629

Calibration of Refueling Water Storage Tank (RWST) level

9/20/13

01374939

Calibration of Refueling Water Storage Tank (RWST) level

2/28/12

00915331

Minor Leakage from 0A WW Pump Well Head

8/20/08

00768385

0B WW PP 10 Year Rebuild

11/09/06

01754077

Received 0A SX Make Up Pp Trouble alarm

7/17/14

00921203

SXCT Fan Assembly Replacement EC 356417

8/23/12

00921198

SXCT Fan Assembly Replacement EC 356417

1/10/07

01634644

Replace Start Contactor Relay K1B at 0SX02PA-B

4/17/13

01682260

Support Diver Insp/Cleaning SXCT South 0B Basin

10/31/14

01691008

Support Diver Insp/Cleaning SXCT South 0A Basin

11/14/14

19

LIST OF ACRONYMS USED

CDF

Delta Core Damage Frequency

AC

Alternating Current

ACIT

Action Tracking Item

ADAMS

Agencywide Document Access Management System

CA

Corrective Action Tracking Item

CAP

Corrective Action Program

CAQ

Condition Adverse to Quality

CCDP

Conditional Core Damage Probability

CDBI

Component Design Bases Inspection

CFR

Code of Federal Regulations

CNMT

Containment

CS

Containment Spray

CV

Chemical and Volume Control

DBA

Design Basis Accident

DC

Direct Current

DRP

Division of Reactor Projects

DRS

Division of Reactor Safety

EC

Engineering Change

ECCS

Emergency Core Cooling System

EOP

Emergency Operating Procedure

ERG

Emergency Response Guideline

FSAR

Final Safety Analysis Report

gpm

Gallons per Minute

IMC

Inspection Manual Chapter

IN

Information Notice

IR

Inspection Report

LCO

Limiting Condition for Operation

LERF

Large Early Release Frequency

LLC

Limited Liability Corporation

LOCA

Loss of Coolant Accident

LOOP

Loss of Offsite Power

LTOP

Low Temperature Overpressure Protection

MCC

Motor Control Center

MOV

Motor-Operated Valve

NCV

Non-Cited Violation

NEI

Nuclear Energy Institute

NOV

Notice of Violation

NPSH

Net Positive Suction Head

NRC

U.S. Nuclear Regulatory Commission

NRR

Nuclear Reactor Regulation

PARS

Publicly Available Records System

PM

Preventive Maintenance

PORV

Power-Operated Relief Valve

PRA

Probabilistic Risk Assessment

RASP

Risk Assessment Standardization Project

RCS

Reactor Coolant System

RHR

Residual Heat Removal

RIS

Regulatory Issue Summary

RWST

Refueling Water Storage Tank

20

SAPHIRE

Systems Analysis Programs for Hands-on Integrated Reliability Evaluations

SDP

Significance Determination Process

SI

Safety Injection

SPAR

Standardized Plant Analysis Risk

SR

Surveillance Requirement

SRA

Senior Reactor Analyst

SSC

System, Structure, and Component

SSDPC

Safety Systems Design, Performance and Capability Inspection

SX

Emergency Service Water

SXCT

Emergency Service Water Cooling Tower

TORMIS

Tornado Missile Risk Evaluation Methodology

TS

Technical Specification

UFSAR

Updated Final Safety Analysis Report

UHS

Ultimate Heat Sink

URI

Unresolved Item

VAC

Volts Alternating Current

VDC

Volts Direct Current

WOG

Westinghouse Owners Group

B. Hanson

-3-

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public

Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy

of this letter, its enclosure, and your response (if any) will be available electronically for public

inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS)

component of the NRC's Agencywide Documents Access and Management System (ADAMS).

ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html

(the Public Electronic Reading Room).

Sincerely,

/RA/

Christine A. Lipa, Chief

Engineering Branch 2

Division of Reactor Safety

Docket Nos. 50-454; 50-455

License Nos. NPF-37; NPF-66

Enclosures:

(1) Notice of Violation

(2) IR 05000454/2015008; 05000455/2015008;

cc w/encl: Distribution via LISTSERV

DISTRIBUTION w/encl:

Kimyata MorganButler

RidsNrrDorlLpl3-2 Resource

RidsNrrPMByron Resource

RidsNrrDirsIrib Resource

Cynthia Pederson

Darrell Roberts

Richard Skokowski

Allan Barker

Carole Ariano

Linda Linn

DRPIII

DRSIII

Jim Clay

Carmen Olteanu

ROPreports.Resource@nrc.gov

ADAMS Accession Number ML15203A042

Publicly Available

Non-Publicly Available

Sensitive

Non-Sensitive

To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy

OFFICE

RIII

RIII

RIII

RIII

NAME

MJones for NFeliz-Adorno:cl

CLipa:

DATE

07/21/15

07/21/15

OFFICIAL RECORD COPY